MANUAL FOR AEC-NSF INSTITUTE RADIOISOTOPE TECHNOLOGY byLloyd E. Brownell June, 1961

g C-b t

PREFACE Chapters 1-5 of this manuscript were originally written for possible use in the text "Radiation Uses in Industry and Science", prepared for AEC by L. E. Brownell and published by the US Government Printing Office, June, 1961. Chapters 1 and 2 of this manuscript were originally combined for use in the above mentioned text but were not used because revisions and up-dating to include 1961 maximum permissible radiation doses could not be completed in sufficient time to meet the publication schedule. Chapters 3-5 of this manuscript were not included in the above mentioned text because the subject material in these chapters was considered to be too technical. These first five chapters mentioned above have' been revised and printed at the University of Michigan for special use in teaching the AEC-NSF 1961 Summer Institute in Radioisotope Technology and as a supplement to the text "Radiation Uses in Industry and Science". The remaining chapters are based on the laboratory work conducted at the University of Michigan during the Summer Institute. The author wishes to acknowledge the extensive assistance received by many co-workers in preparing the material in this manuscript. Dr. John Nehemias, former health physicist at the Fission Products Laboratory, and Dr. Ardath Emmons, former Supervisor of the Michigan Memorial Phoenix Project Laboratories, were major contributors to the original version of Chapters 1 and 2, respectively. Chapter 1 was reviewed and up-dated with assistance from Messrs. Lowell Yemin and John Jones, Health Physicists, Radiation Control Service, University of Michigan. Chapters 3 and 4 were revised by Messrs. N.E. Kothary and J.E. Sickles, respectively, of the Graduate School. Chapter 5 was prepared using lecture notes of Professor W. Kerr, Department of Nuclear Engineering and class notes of J.V. Nehemias. Chapter 6, "Experimental Techniques in Nuclear Tracer Studies" was prepared by Professor Adon Gordus, of the Department of Chemistry, the University of Michigan. Chapter 7, "Experiments in Radioisotope Technology" was prepared by Professor Lloyd Brownell and Messrs. R. Borcherts and J.Sickles of the Department of Nuclear Engineering, the University of Michigan. Chapter 8 "Nuclear Radiation Detection and Measurement", was prepared by Professor Geza Gyorey and Mr. Philip Pluto also of the Department of Nuclear Engineering, the University of Michigan. The reader should recognize that these three chapters are in major part a reproduction of class notes and were not prepared using the same format used in the first five chapters. Thanks is given to many other individuals who helped in various ways and to authors and publishers for permission to reproduce tables and figures used in this manuscript. L. E, Brownell Professor of Nuclear and June 16, 1961 Chemical Engineering

TABLE OF CONTENTS Chapters Page I. Safety in Work With Radioisotopes 1.1 Early history of radiation exposure 1.1 1.2 Recent developments 1.2 1.3 Safety in the atomic energy industry 1.2 1.4 Health Physics 1.6 1.5 Radiological Units 1.10 1.6 Biological effects of radiation 1.14 1.7 Radiation syndrome 1.14 1.8 Effect on cancer incidence 1.16 1.9 Effect of life span 1.16 1.10 Radiosensitivity of different tissues 1.17 1.11 Effect of rate of irradiation 1.18 1.12 The influence of the nature of the radiation 1.18 1.13 Internal hazard 1.20 1.14 Possible effect of radiation on subsequent generations 1.20 1.15 Shortening of life span 1.22 1.16 Significance of radiation effects 1.22 1.17 Maximum permissible exposure levels 1.22 1.18 Accumulated dose (radiation workers) 1.25 1.19 Emergency dose (radiation workers) 1.26 1.20 Medical dose (radiation workers) 1.26 1.21 Dose to persons outside of controlled areas 1.26 1.22 Operational and administrative guides 1.27 1.23 Regulation and control by AEC 1.36 1.24 Differences between AEC regulations and ICRP recommendations 1.37 1025 Practical limits-comparison with "background" 1.38 1.26 Comparison with nonoccupational exposures 1.40 1.27 Three safety rules for protection from external radiation 1.42 II. The Design and Use of Radiation Laboratories 2.1 Hot cells 2 3 2.2 Manipulators 2.7 2.3 Viewing techniques 2.8 2.4 Hot-cell operational problems 2.10 2.5 Junior caves and shielded glove boxes 2.12 2.6 The low-level or radioisotope laboratory 2.14 2.7 Common hazard parameters 2.19 2.8 Exhaust-air control 2.19 ii

TABLE OF CONTENTS (CON'T) Chapters Page II. Con't. 2.9 Monitoring with instruments 2.20 2.10 Personnel film badges 2.21 2.11 Use of special clothing 2.24 2.12 Decontamination procedures 2.25 2.13 Radioactive wastes 2.32 2.14 Example design of a multi-purpose hot lab 2.35 III.Film, Glass, Chemical and Calorimetric Dosimetry 3.1 Film Dosimeters 3.1 3.2 Film Badges 3.1 3.3 Polyvinyl Chloride Films 3.9 3.4 Cellophane Films 3.14 3.5 Coloration in Plastics 3.17 3.6 Chemical Changes in Plastics 3.17 3.7 Glass Dosimetets 3.19 3.8 Phosphate Glass Personnel Dosimeter 3.19 3.9 High-Dosage Phosphate Glass 3.25 3.10 High Dosage Cobalt Glass 3 31 3.11 Chemical Dosimeters 3.31 3.12 Ferrous-Ferric 3 35 3-13 Ferrous Sulfate-Cupric Sulfate 3.40 3.14 Ceric Sulfate 3.41 3.15 Shlorinated Hydrocarbons 3.43 3.16 Gaseous Nitrous Oxide 3.44 3.17 Other Chemical Systems 3.45 3.18 Calorimetric Dosimeters 3 47 3.19 Luminiscence Degradation 3.49 3 o20 Microbial Monitors 3 ~52 3.21 Summary 3.52 IV. Gamma Shielding 4.1 Attenuation of Gamma-Radiation from Point Sources 4.1 4.2 Narrow Beam Attenuation 4.5 4.3 Half-Value and Tenth-Value Thicknesses 4.5 4.4 Mixed Energies 4.11 4.5 Attenuation "Build-up" Factors for Point Sources 4.11 4.6 "Broad-Beam" Coefficients 4.15 4.7 Energy Absorption Coefficients, ua 4.15 4.8 Example Problem 1, Point Source 4.20 4.9 Calculation Procedures for Various Geometries and Multiple Shields 4.24 4.10 One-Material Shield 4.26 iii

TABLE OF CONTENTS (CON ' T) Chapters Page IV. Con't. 4.11 Several Slabs of Different Materials 4.31 4.12 Effect of Geometry (from Rockwell) (13) 4.32 A. Point Source 4 33 B. Line Source 4.34 C. Disk Source (K curves) 4. 36 4.13 Example Problem 2, Point Source Using Method from Rockwell(l3) 4.38 4.14 Example Probltm 3, Line Source Using Method from Rockwell[13 4.39 4.15 Infinite Slab Source(4) 4.46 4.16 Cylindrical Source(4) 4.53 4 17 Spherical Source(13) 4.66 4.18 Types of Gamma Radiation Sources 4,72 4.19 MTR Fuel Elements 4.72 4.20 Example of Gamma Shielding Calculation for MTR Fuel Elements 4.73 V. Counting Nuclear Radiation 5.1 Auxiliary and counting equipment 5.1 5.2 The counting of randomly occurring events 5.4 5.3 The Poisson distribution 5.5 5.4 Mean value 5.8 5.5 Coincidence 5o10 5.6 Deviations from the mean value 5.11 5.7 The effect of background on the interpretation of counting data 5.15 5.8 The standard deviation 5.18 5.9 Percentage probable error 5.21 5.10 Corrections in beta counting 5.21 5.11 Dead time 5.23 5.12 Geometry 5.26 5.13 Tube efficiency 5.29 5.14 Process efficiency 5.31 5.15 Forescatter 5.31 5.16 Backscatter 5.33 5.17 Self-scatter 5 35 5.18 Self-absorption 5.35 5.19 Window and air absorption 5.37 5.20 Energy 5.39 5.21 Relative importance of correction factors in beta counting 5.40 5.22 The counting of gamma emitters 5.42 iv

TABLE OF CONTENTS (CON'T) Chapters Page V. Con't. 5.23 Counting alpha particles 5.45 VI. Experimental Techniques in Nuclear Tracer Studies 6.1 Introduction (By L.E. Brownell) 6.1 6.2 Outline for first three weeks of study (AoA. Gordus) 6.3 6.3 General Laboratory Rules for Nuclear Chemistry 6.7 6.4 Scale of 64 6.8 6,5 Experiment 1 Counter Operation (Adapted from G. Wilkinson) 6.10 6,6 Experiment 2 Backscattering 6.17 6.7 Experiment 3 (Adapted from Go Wilkinson) 6.19 6.8 Experiment 4 Radiochemical Separations - ParentDaughter Decay Separation of Ce144 from Pr144 by Precipitation 6.27 6.9 Experiment 5 Isotope Dilution 6.31 6.10 Determination of Percent Halogen by Activation Analysis 6.33 6.11 Nuclear Properties of Selected Isotopes 6.37 6.12 Observed Activity vs. Time 6.38 6.13 Radioactive Decay D.Ec Hull 6.39 6.14 Fractional Midpoint of Count vs. Duration of Count 6.40 6.15 Selected Charts of Activity vs. Time 6.41 VII.Experiments in Radioisotope Technology 7.1 Experiment 1 Measurement of Radiation Field of Small and Large Cobalt-60 Sources Using Various Instruments Operated by Ionization of Gases 7.1.1 7.2 Experiment 2 A chemical dosimetry and dye (Film) dosimetry 7.2.1 7.3 Experiment 3 "Broad Beam" attenuation of gamma radiation 7.3 1 7.4 Experiment 4 Radiography with a source-target mixture 7.4.1 7.5 Experiment 5 Area Decontamination 7.5.1 7.6 Experiment 6 Treatment of radioactive wastes 7.6.1 7.7 Experiment 7 Measurement of Per Cent moisture by neutron slowing down 7.7.1 7.8 Experiment 8 Germination and growth of irradiated seeds and sprout inhibition in tubes 7,8.1 7,9 Experiment 9 Radiation of food 7.9.1 7.10 Experiment 10 Effects of radiation on chemical reactions 7.10.1 v[

TABLE OF CONTENTS (CON'T) Chapters Page VII.Con't. 7.11 Experiment 11 Glass dosimetry-cobalt type 7.11.1 VIII.Nuclear Radiation Detection and Measurement 801 Discussion 8.1 8.2 General laboratory procedures 8.1 8.3 List of recommended equipment 8.2 8.4 Introduction to theory of instrumentation 8.5 8.5 Derivation of charged particle energy matter 8.8 8.6 Ionization chambers 8.10 8.7 Mean level chamber 8.11 8.8 Pulse ionization chambers 8.14 8.9 Input voltage pulse forms to a circuit caused by the production of ion pairs between two electrodes 8.17 8.10 Experiment on the use of ionization chambers 8.22 8.11 Proportional chambers 8.23 8.12 A note on gas multiplication 8.27 8.13 Experiment on the use of the proportional counter 8.29 8.14 Geiger-Muller chamber 8.30 8.15 Experiment on the use of the GM chamber 8.34 8.16 Statistics 8.35 8.17 Poisson distribution 8.35 8.18 Interval distribution 8.38 8.19 Checking equipment using count rate data 8.38 8.20 Degrees of freedom 8.41 8.21 Test of fit of sample to poisson 8.41 8.22 Several simultaneous statistical fluctuations 8.41 8.23 Counting loss due to chamber dead time 8.43 8.24 Experiment on the statistics of counting 8.44 8.25 Scintillation detectors 8.46 8.26 Scintillation mechanism 8.47 8.27 Analysis of gamma ray pulse distribution (Jointly with W. Smith) 8.50 8.28 Details on the conversion of gamma energy to a measureable pulse from the phototube (Jointly with J. Trombka) 8.56 8.29 Conversion of light energy to photo-electrons and subsequent multiplication 8.57 8.30 Pulse height vs. energy 8.59 8.31 Energy resolution 8.59 8.32 Detection efficiencies 8.62 8.33 Experiment using a scintillation counter 8.64 vi

TABLE OF CONTENTS (CON'T) Chapters Page VIIIo Con't. 8.34 Experiment on scintillation spectrometry 8.65 8.35 Neutron detection 8.67 8.36 Experiment on neutron detection using BF3 tubes 8.68 8.37 Thermal neutron flux measurements in a nuclear reactor using activation techniques 8.70 8.38 Experiment on neutron detection by induced activity 8.74 8.39 Fundamentals of junction-type solid state ionization detector (Jointly with G.Brown) 8.78 8.40 Junction detector 8.78 8.41 Fermi statistics 8.79 8.42 Effect of impurity concentration change and bias on the fermi level 8.81 8.43 Depletion depth and barrier capacitance 8.82 8.44 Balance equation and pulse output 8.86 8.45 Range of particle in silicon 8.87 8.46 Energy loss by incident -particle 8.88 8.47 Limiting energy of ionization (Ei) 8.88 8.48 Time required for to complete ionization 8.90 8.49 Estimation of charge collection time and charge collected 8.90 8.50 Estimation of voltage output of the detector 8.92 8.51 Experiment on alpha particle detection using the P-N junction 8.94 8.52 Questions 8.96 8.53 Ionization chambers 8.96 8.54 GM-Proportional counters 8.97 8.55 Statistics 8.98 8.56 Scintillation spectrometry 8.99 8.57 Neutron detection 8.100 8.58 Foils 8.101 8.59 Solid state detector 8.101 8.60 General 8.102 8.61 Calculational problems 8.104 vii

LIST OF FIGURE CAPTIONS Chapters Page I. Safety in Work With Radioisotopes 1.1 Withdrawing irradiated material from reactor core 1.8 1.2 Filter type masks and protective clothing 1.8 1.3 High level radiation area at the Hanford Atomic Products 1.9 1.4 Pencils and badges worn to obtain records of radiation exposure 1.9 1.5 Effect of different types of radiation on tissue 1.12 1.6 Typical "S" Curve for radiation-induced death with L-D 50 of 450 roentgens (for man) 1.12 1.7 Exposure from common X-ray examinations 1.41 1o8 Engire crew moving car loaded with highly radioactive waste to Hanford burial site. 1.41 1.9 Burial of highly radioactive waste at Hanford 1.43 1.10 Operator working behing heavy shield and using remote manipulator 1.43 II. The Design and Use of Radiation Laboratories 2.1 "Hot Cells" 2.4 2.2 Hot cells in the Michigan Memorial-Phoenix Laboratory constructed of barytes concrete in a steel-plate shell and equipped with Argonne Model 8 manipulators 2.4 2.3 Canyon view in radiometallurgy building at Hanford showing battery of hot cells constructed of cast iron 2.4 2.4 Ball-socket manipulator in radiation-shield wall 2.4 2.5 Argonne Model-8 Manipulator, illustrating the degrees of freedom 2.8 2.6 Sectional diagram of a periscope for viewing through the shield wall 2.8 2.7 Isometric sketch showing construction details of a zinc-bromide window 2.8 2.8 Isometric sketch showing construction details of a lead-glass shield.window 2.8 2.9 Junior cave used at Hanford 2.12 2.10 On the left, shielded-box assembly with ballsocket manipulators. On the right, standard glovedbox unit 2.12 211 Interior view of a shielded-box unit equipped for remote radiochemistry work 2.12 2.12 Photograph of a radiochemistry laboratory, showing the hood installations and gloved-box manifold assembly on the right. 2o12 viii

LIST OF FIGURE CAPTIONS (CON'T) Chapters Page II. Con't. 2o13 Photograph of a "'walk-in" hood capable of housing assemblies 6-1/2 feet high. A standard chemistry hood-bench assembly is visible in the background., 2.14 2,14 Plan view of typical one-room radioisotope tracer laboratory 2.14 2.15 Isometric view of typical one-room radioisotope tracer laboratory 2.14 2.16 Isometric and plan view for a typical threeroom radioisotope tracer laboratory 2l14 2.17 Floor plan of a radiochemical laboratory building with radiation zoning and airflow planning 2.20 2.18 Double hood installation with exhaust-air filtration boxes above. 2 20 2.19 A fixed ionization chamber which measures radiation level at point inside shielding wallo 2.20 2.20 Routine check of hands and feet of workers for possible contamination before leaving building at Hanford Plant 2. 20 2.21 Worker with special suit for protection in radiocontaminated areas being checked for contamination 2 ~24 2,22 Routine check of floor for possible contamination 2,24 2,.23 Radioactive liquid-waste tank farm of 10,000 -gallon storage capacity, 2.24 2o24 Michigan Memorial-Phoenix Laboratory hot-lab floor plan 2036 III. Film, Glass, Chemical and Calorimetric Dosimetry 3o1 Insertion of film in film badge 3e3 3.2 Range and sensitivity of typical film 3~4 3.3 Relative sensitivity vs, effective energy for duPont 502 emulsion and filtered X-radiation 3.6 3.4 Cross-section of X-Ray and beta-gamma film badges 3.5 Cross section of neutron film badge 3.8 3.6 Percentage transmission of PVC film vs. recprocal of relative dosage from Cobalt-60 gammas 3.13 3a7 Percent transmission vs. gamma-radiation dosage for cellophane film 3015 3o8 Change in percent transmission vso gammasradiation dosage for cellophane film 3.16 309 Comparison of gamma photon and electron radiation and different dose rates for cellophane films 3,18 ix

LIST OF FIGURE CAPTIONS (CON'T) Chapters Page III. Con'te 3o10 Calibration curve for electron radiation of cellophane films 3 21 3.11 Spectra of nonirradiated (Ag centers) and irradiated (Ag~) phosphate glass 3,22 3.12 DT-60/PD Dosimeter response 3~23 3.13 Phosphate glass dosimeter, (A) assembled topside, (B) assembled underside, (C) cover portion showing lead shield, (D) base portion showing glass block, and (E) wrench 3~24 3.14 One type of reader for phosphate glass dosimeter 3.26 3.15 Photograph of phosphate glass blocks before and after irradiation 3,27 3.16 Absorption spectra of phosphate glasses 3.28 3.17 Dose dependence of absorption of phosphate glass measured at different wavelengths 3,29 3.18 Dependence of phosphate glass sensitivity upon energy of radiation 3,30 3.19 Stabilization of coloration of silver-activated phosphate glass by thermal acceleration of fading 3, 32 3.20 Photograph of the small-volume phosphate dosimeter 3e33 3.21 Micromoles of ferrous ion oxidized as a function of the total irradiation 3 36 3.22 Micromoles of ceric ion reduced as a function of the total irradiation 3.37 3.23 Calibration and conversion curve for Beckman Model DU Meter in the Fission Products Laboratory, the University of Michigan 3 39 3c.24 Comparison of response curves for Fricke and Ferrous-Cupric dosimeter 3 42 3.25 Calibration curve for N20 dosimeter 3,46 3.26 Construction of calorimeter cylinder 3 50 3,27 Horizontal cross section showing calorimeter construction 3. 51 IV. Gamma Shielding 4.1 Flux reduction with distance 4,3 4.,2 Good-geometry configuration 4 o 4 4.3 Bad-geometry configuration 4. 4 4,4 Linear attenuation coefficient of gamma rays in lead 4. 6 4 5 Total mass attenuation coefficient for X- and gamma radiation in the range of photoelectric effect 4.7 4 6 Total mass attenuation coefficients for X- and gamma radiation in the range of Compton scatter 4.8 x

LIST OF FUGURE CAPTIONS (CON'T) Chapters Page IV. Con't.o 4~7 Total mass attenuation coefficients for X- and gamma radiation in the range of pair formation 4.9 4,8 Narrow beam tenth value thicknesses of various materials for gamma radiation 4.12 4,9 Diagram of "build-up" in lead 4 13 4,.10 "Broad beam" transmission of radium, cobalts60 and cesium-137 gamma rays in concrete 4 Q16 4.11 "Broad beam" transmission of radium, cobalt-60 and cesium-137 gamma rays in iron 4o17 4,12 "Broad beam" transmission of radium, cobalt-60 and cesium-137 gamma rays in lead 4.18 4.13 Specific radiation flux,, as a function of energy in Mev of gamma radiation 4.19 4.14 Configuration for example problem 1. 4 20 4.15a Dose build-up factor in lead for isotropic point source 4.27 4.15b Energy-absorption build-up factor in lead for isotropic point source. 4.27 4,16a Dose build-up factor in iron for isotropic point source. 4.28 4.16b Energy-absorption build-up factor in iron for isotropic point source e 4.28 V. Counting Nuclear Radiation 5,l Vertical lead shield 5 3 5.2 Scaler 5.3 5.3 Time Interval Divided into k units each l/k wide 5k6 5~4 The standard error in psrticle counting 5 22 5.5 Dependence of error on counts 5 22 5.6 Percent probable error vse total number of counts 5,24 5.7 Considerations in radioactivity measurements 5.24 5,8 Dead time in a GM counter 5.25 5.9 Schematic diagram of coincidence losses due to dead time 5.25 5o10 Geometry for source of radius " c" 5 28 5.11 Geometry for source of radius nci" 5.28 5 12 Effect of asymmetry on efficiency of counting 5 32 5o13 Percent forescattering as determined with polystyrene 5 32 5o14 Backscatter as a function of atomic number and energy of radiation 5.34 5o15 Saturation backscatter 5 o34 xi

LIST OF FIGURE CAPTIONS (CON'T) Chapter Page V. Con' t 5,.16 Self-absorption 5 38 5.17 Absorption in air and window 5.38 5.18 Effect of energy on beta counting 5.41 5.19i Effect of sample w6ight on self-absorption coefficient in beta counting 5.41 5.20 Counting rates, counts minute 5.44 VI. Experimental Techniques in Nuclear Tracer Studies 6.1 'Diagram of counter 1rangement 6.18 6.2 Count rate with Ce~ Predominating 6.30 6.3 Count rate with Prl44 Predominating 6.30 6.4 Fractional midpoint of count vs. duration of count 6.40 6.5 Activity vs. Time 6.41 6.6 Activity vs. Time 6.41 6.7 Activity vs. Time 6.41 6.8 Activity vs. Time 6.41 6.9 Activity vs. Time 6.41 6.10 Activity vs. Time 6.41 VII. Experiments in Radioisotope Technology 7.1 Small cobalt-60 source 7.02a 7.2 Cutaway perspective view of radiation curve 7.02a 7.3 Plan and elevation sectional view of radiation cave 7.02b 7.4 Door interlock to radiation cave 7.02c 7.5 Loading cobalt-60 rods into holder under 16 feet of water used as shielding 7.02c 7.6 Original radiation flux measurements made on small cobalt-60 source 7.04a 7.7 Dose rate on midplane of 10- c Co60 source 7.04b 7.8 Dose rate on axis of l-kc Co source 7.04b 7.10 Calibration and conversion Curve for Beckman Model DU meter in FPL 7.2.1a 7.11 Sketch showing general shape of curve for maximum Bragg-Gray effect as determined by blue cellophane dosimetry 7.2.3 7.12 Experimental observations of the change of pH of freshly prepared chloral hydrate solutioA vs. dosage of gamma radiation 7.3 la 7.13 Absorption measurements performed in the radiation '!cave" at the Fission Products Laboratory 7.3-3 xii

LIST OF FIGURE CAPTIONS (CON'T) Chapter Page VII. Con't. 7.14 Comparison of the spectra of three radiographic sources 7.4.3a 7.15 Radiograph of human hand, 79 hr. exposure at 20 in. from a promethium tungstate source 7.4.3b 7,16 Technical radiograph, 44 hr. exposure at 10 in. from a promethium tungstate source 7.4.3b 7.17 Checking shoe covers at the Hanford Plant for contamination before stepping onto stepoff pad. 7.5.2a 7..18 Periscope picture of equipment in decontamination canyon at Hanford Plant 7.5.2a 7.,19 A Hanford Works "canyon" building over 800 feet long 7.5.3a 7.20 Typical reactor area at Hanford 7.6.la 7.21 Photographic inspection of underground tank used for storage of waste fission products at Hanford 7 6.lb 7.24 Plot of growing radishes from irradiated seeds 7.804a 7.25 Typical radish plants from irradiated seeds 7.8.4a 7.26 Irradiated and nonirradiated potatoes 7.8.4b 7.27 Irradiated and nonirradiated onions 7.8.4b 7.28 Sample of irradiated and nonirradiated smoked salmon after removal from sealed polyethylene bags 7.9.1a 7.29 Irradiated smoked salmon data 7.9. la 7.30 Growth of microorganisms in irradiated meat 7.9.3a 79 31 Irradiated and ~nonirradiated grapefruit after development of mature larvae stage 7.9.3a 7.32 Drawing and specifications for pyrex heavy walled glass reactors 7.10.4 7.33 Drawing of the stainless steel reactor no. E used in the room temperature runs 7.10.4 VIIIoNuclear Radiation Detection and Measurement 8.1 Principal gamma ray interactions 8.7 8.2 Geometry for energy loss derivation 8.8 8.3 Ionization current 8.12 8.4 Typical ionization survey meter 8.13 8.5. The Lauritsen electroscope 8. 13 8.6 Pocket dosimeter 8.13 8.7 Diagram for dharged particles in ionization chamber 8.15 8.8 Charged particle between two electrodes 8.20 8.9 Simplified representation of gas multiplication 8.25 xiii

LIST OF FIGURE CAPTIONS (CONT) Cha pter Page VIIoI Con'to 8,10 A gas flow proportional counter 8,25 8,11 Proportional counter characeristic curves for an Alpha, and a Beta or Gamma emitter 8.28 8.12 Basic circuit for Geiger Muller chamber 8o31 8.13 Need for quenching in a GM 8o31 8o14 Action of quenching gas in GM tube 8.33 8.15 GM characteristic plateau curves 8.33 8,16 Poisson distribution 8.35 8.17 The normal distribution 8.37 8.18 Representative block diagram for scintillation counter 8.51 8.19 Scintillation counter 8051 8o20 Detail on two types of photo-multiplier tubes 8.52 8o21 Typical well counter (Courtesy RCL Inc.) 8.52 8.22 Idealized differential curve for monoenergetic gammas 8 53 8.23 Conversion efficiency Cnp ( ) and Spectral sensitivity S ( ) of type S-ll photocathodes 8.57 8.24 Typical source-crystal geometry 8.62 8 25 '8.26 Typical neutron absorption cross section, a US energy - 8o 70 8,27 Neutron flux, U.oS energy 8 70 8028 Idealized absorption coefficient vs. energy for screening slow from fast neutrons 8,71 8.29 Approximate absorption coefficient for cadmium 8o72 8.30 Detector foil activity during irradiation and counting 8 76 8.31 Detection foil activity vso cadmium cover thickness 8.77 8.32 Detector activity per unit thickness vs. detection thickness 8.77 8.33 Relative count rate vs. edial 8.95 8.34 Percent resolution vs. detector bias 8,95 8.35 Pulse height at E vs. detector bias 8,95

.Chapter 1 Safety in Work With Radioisotopes The problem of safety when working ionizing radiation is unique in that in almost all cases the htuaa senses provide no warning of dangerous levels of radiation, Safety in work with radioactive material requires an understanding.of potential hazards associated with its use plus an understanding of the methods of evaluating the hazard. This chapter briefly discusses qualitatively the problem of radiation hazards and the biological effects of radiation exposure. The quantitative aspects of the problem, units of radiation.dosage, tolerances, permissible exposure levels, and safety rules are then considered. 1.1 Early history of radiation exposure The animal body has a variety of defenses. against most natural dangers, b-ut it has no means of instantaneously detecting ionizing radiation which cannot be seen, felt, heard, smelled or tasted. - The odor of ozone may be detected in air exposed to inatense ionizing radiation, but the radiation level required to produce this detectable quantity-of ozone is tremendously higher than minimum danger levels. Thus it is necessary to depend on instruments for the-detection of radiation., The early -history of work with.-radiation is marred by many cases.of serious overexposure to external radiation (l-6) Within -thirty days of the discovery of X-rays by Roentgen in 1896, one of his co-workers reported hand dermatitis so acute that medical aid was required (4). Another radiation injury.occurred to Henri Becqyerel (5).: Several years after his discovery of.radioactivity,. he carried a glass tube containing a radium compound in his vest pocket for six days.. An ugly red ulcer developed on his body, which healed in a.few months but which left a permanen't scar. Pierre Curie voluntarily exposed his. arm for several: hours to the action of radium. A lesion resulted, resembling a burn, which required several months to heal..(6). In the early days, the hazards of radiation exposure were not considered important by public health authorities-. Relatively few workers were -ever exposed to ionizing radiation and these were- considered to be highly com — petent: skilled technicians. In spite of.early accidents resulting from faulty equipment and ignorance of the hazards involved, radiation control was left largely to the individual radiation worker (78). 1-1

1-2 During the first World War the use of paints containing traces of radium f or dial illumination and related applications became widespread. No radiation protection measures were observed. Some of the dial painters habitually pointed their brushes on their tongues. A number of the women employed in this industry have developed bone sarcoma and other radiation injuries. The first clinical symptons exhibited by these women appeared from 10 to 30 years after exposure to radium handling had terminated (9,10). 1.2 Recent developments The development of the X-ray machine with its world-wide installation in medical and dental centers, and the advent of nuclear energy with the widespread use of many types of radiation, have increased the hazards of radiation injuries by many orders of magnitude. Yet the serious overexposures that have occurred since December 1942 (when the first nuclear reactor was completed) have involved a much smaller percentage of those working with radiation. One lethal injury occurred when a scientist studying criticality of a fissionable mass was obliged to reach in and quickly separate reacting components in order to avoid a catastrophe (11). An operating nuclear reactor presents a unique combination of various radiation hazards. The reacting core is an intense cource of neutron, gamma, and beta radiation. The beta-particles are absorbed in the reactor core but neutrons and gamma phatons are very penetrating. Personnel working in the vicinity of the reactor must be protected against these latter two types of radiation. Because different shielding materials provide optimum protection for these two types of radiation, reactor shielding calculations and design must be twofold. A further hazard is presented by the radioactivity induced in any material,. experimental., structural, or otherwise, which has been exposed to the neutron flux at or near the core. Any material removed from the reactor is itself an additional radiation hazard and must be treated as such. Also, as fission proceeds, fission products are formed which are intensely radioactive. Removal, manipulation, and reprocessing of spent reactor fuel elements must be performed under strict radiation control. Figure lo. shows an irradiated sample being removed from a reactor into a portable lead tunnel, in which it can be moved safely to the site of the experiment. The radiation level is being read and recorded. 1.3 Safety in the atomic energy industry The remarkably small number of radiation injuries occurring in the atomic energy program to date demonstrates the efficiency of safety procedures in protecting personnel of this new industry from the unique hazards of radiation. During the operation of the Manhattan Project from 1942 to 1946 there were only two accidents that involved radiation injury to personnel. The same high standard of safety has been maintained since the transfer of this project to the AEC. Tables 1.1, 1.2, and 1.3 summarize the experience of the 32 major AEC operating contractors in routine operationo- In addition, between August.L, 1945, and July 3, 1956, there were 16 radiation accidents in which 69 persons were overexposed of which two died and 19 suffered radiation burns (12).

Table 1,1 shows that the average exposure per worker was only a small percentage of the yearly maximum permissible dose. From 1947 to 1955 only 0.34% received over 5 rem (roentgen equivalent man) in routine operations. The permissible dose during this period,.was 15 rem/year (the present permissible dose is 5 rem/year). This attention to radiation safety has also resulted in a remarkably low rate of accidents of all types and the overall safety records of plants in the atomic energy program are considerably better than those of the average industrial plant. The atomic energy industry is one of the safest industries in which to work, as is indicated by Table 1.4 (13). Table 1.1 Exposure of Personnel Working for AEC Contractors (12)(*) External Radiation - 1947-l955 151955 only 1 958 only Annual j Number Per Cent Number I; Per Cent'| Number I Per Cent Total of j of of of of. I of Amount J Workers L Total Workers I Total, | Workersl Total Q-lrem J 186,836 95-34 J 56,708 f 94.21 659 316 j 90 -1-5 8.468 4.32 3,157 5.24 6393 f 9.7 5-15 642 0.33 f 326 f 0.54 181 002 F I I 7-15 19 3 0005 1 0.01.018 I B.... / I 1 1 (*) Major Activities in the Atomic Energy Programs Jan.-Dec. 1959. USAECG Jan. 1960

Table 1.2 Highest Accumulated Yearly Exposure in Rem. to Individual Employees of AEC Contractors During Routine Operations (12) (Does not include accidents) 1947 1948' 1949 1950 1951 1952 1953 1954 1955 lest Dose 23.5 20.3 13.6 9.0 7.1 15.7 12.9 27.8 17.9.,age of Highest Doses (a) 7.4 7.8 4.0 3.9 2.8 5.0 9 6.5 5.8.age of 10 Highest Doses (b) 5.2 4.2 2.6 2.2 1.8 2.9 3.4 3,9 4 Indicates average of single highest exposures which occurred each year. Indicates average of 10 highest annual exposures.

Table 1.3 Fraction of Permissible Body Burden of Aec Contractor Employees by Isotope. Jaruary 1947 - June 1956o Figures are for Maximum values obtained for each person in each year (12). Percent of Uran- Fission Pluto-' Polo- Stron- Fission 'i_::Tri- Ruth- Ces- AmerPermissible ium Prod'ts nium nium tium Prod'ts tium enium ium icium Total Body Burden (a) + Other Isotopes Less than' 10 1 i4,481 10,037 6,394 4,584 4,537 1,598 102 28 5 0 41,766 10-50 15, 564 3 25 475 47 1 8 O0 0 0 16,123 50-100: 8,177' 0 5 129 9 0 1 0 0 2 8,323 More than 100 4,844 0 5 59 1 0 1 0 0 0 4, 910 (a) Based on chemical toxicity of uranium which is greater than its radiation hazard.

1-6 Table 1.4 Injuries per Million Man Hours (13) 1948 1949 Lumbering Industry 49.04 47.72 Air Transport 15.05 12097 Shipbuilding 10.14 8.86 Automobile 8.47 6o 35 Steel 5.86 4.96 All Chemical Industries 7.51 5.72 All Industries 11o49 10 14 Oak Ridge National Laboratory 2.03 1.28 The excellent record of Oak Ridge National Laboratory illustrated in Table 1.4, and of other national laboratories, is the result of careful, competent safety work in all fields and particularly in the new field of "health physics." The Hanford Atomic plant in the state of Washington has had an excellent record with no deaths or permanent injuries as a result of work with radiative materials. Earlier workers in the field of radiation had a-much poorer safety record. 1.4 Health Physics The new field termed "health physics" has appeared as a direct result of the birth and rapid development of the atomic energy industry. Health Physics may be defined as the science and art of radiation protections. It is.a science devoted to the study, evaluation, and control of ionizing radiation. The health physicist's responsibility is the protection of personnel from radiation injury which may result from ignorance of-radiation hazards or from preoccupation with other duties which preclude adequate attention to these hazards. The health physicist must be thoroughly trained in working with radioactive materials and.familiar with the nature of the hazards and the means of measurement and protection. Among his duties are: 1. Protecting personnel from radiation, and, in those instances where some exposure is unavoidable, maintaining exposure levels below the maximum permissible tolerance values set by the Atomic Energy Commission.

1-7 3. Requiring that appropriate protective clothing, such as gloves and respirators, be worn in potentially dangerous radiocontaminated areas. Figure 1.2 shows two laboratory workers at the Hanford Atomic plant wearing protective clothing and filter type masks in a contaminated atmosphere while checking some contaminated waste. 4. Establishing and supervising adequate shielding procedures, decontamination, indicating radiation danger areas, radioactive monitoring of areas, laboratories, and personnel, and maintaining records of radiation and exposure levels. Figure 1.3 shows the entrance to a high-level radiation area at Hanford marked for protection of personnel. In this area specially designed clothing for use in contaminated areas and radiation detection equipment are used 5. Providing surveillance and coordination on over-all aspects of radiation protection and notifying laboratory supervisors immediately of all violations of good health practice. Figure 1.4 shows two types of devices used to measure the cumulative external exposure that an individual may receive while working with sources of ionizing radiation. One of these "pencil" devices is the "pocket chamber" which is actually a pocket sized ionization chamber charged to approximately 150 volts prior to use. An electrometer is needed to read the risidual charge on the chamber after a period of use. The electrometer is calibrated so that the scale reads directly in milliroentgens. A similar "pencil" device, but perhaps more useful in certain instances, is the self reading pocket electroscope. This instrument, in addition to the pocket air condenser chamber contains its own electroscope calibrated in milliroentgens, enabling a person to keep constant check on his cumulated exposure during the course of an experiment. The pocket chamber devices are seldon used as the sole record of a person's exposure because of technical difficulties inherent in this type of device, If a chamber is inadvertantly dropped or bumped against an object, it may be discharged thereby indicating a high reading. Also, a chamber capable of measuring the maximum permissible dose of radiation (to man) will not measure extremely high doses. The badge shown in Figure 1.4 serves both for identification ans as a record of radiation exposure. It contains two films sensitive to X-ray, beta-particle, and gamma radiation. One film is very sensitive to low -levels of radiation and the other is sensitive to high levels. The badge also contains various filters which may be made of copper, lead, or cadmium. These filters permit identification of the degree of penetration of the radiation used in the exposure. Thus, beta-particle radiation may be distinguised from

1-8........~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.. Figure 1.1 Withdrawing irradiated material from reac-or core (Courtesy Oak Ridge National Laboratory (7) _x....:':,.p~~~~f.,5 '2s;;ff'::'fgf::f'Z | | __....................................................................................... lll |....................... _ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~~~~~~~~~~~~~ ~.........n _ | l _ _ _@~~~~~ '-ff '::@::'f_ _ | l __ | i l | " 'v'.''',,.E-f>:~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~:,: fi f~~~~x,.ff e | l | l l |,......................... -, -,,.,-.,.g,,,,.~~~~~~~~~~~~~~~~~.......... ez egfa a WY | l | I I | ai.-.-.E ai.i........................ --- a ~~~~~~~~~~~~~~~~~~~~......... w,.:1Cures of 'X':f;,SSIB;|l * l l | S. <a —E-SSfS ''''' XS | l........................ * I I * '."'' '..................................................... -".'. '. Se'.'.. -'xt.'.S SS' - " S~~~~ ~.......... _ff.S.S~~~~~xs~~fiS XUR7:2s; | l * l l |.,'-.-.............................,' ', -'. '.. -'e-.'. ';~~~~~~~~~~~~~~~~~~~~~~~~~~..............f::,:x, f.& > Rofef~~~~~~~~~~~~~~~~~~~e'. Rgj~~~~~~gXg_ ll l _ lll l l lil _ w-"::i_~~~~~~~~~~~~.......... jfxte~~~~~~~~sexR ishg~~~~~~~~~~~g m-_ _ S _ | l-ghe.. 'exe~ * | I E ___ w _~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.......; x a. _ _ w l _~YfzY~~~~~~~~~~~~~~~~~sf~~~~nf - * l! * _-'s X.>-2. _~~~~~~~~~~~~~~~...... f='':ffY.Re& _ _ | I _i_ | l | __v _ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~...... x I l _s l l | _|- _~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~......... z& hx g2{- a g'le' yfx-::::= yh $@ S l l * l * px,, -,;.-,i-i:,i, i.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~............ m l l l l l | 3,ggxt'.'.',',,, —' '.','.'~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~......: f.R f.: S f,.:: g:::::::.::':f~~~~~~~~~~~~i a i:>'f::R.': f>.-ff:-f ' '' -;:R: fi l l * l *:::::,:: i~~~~~~~~~~~~~~~~~~~~~p.,-' - ' i - g ''- ' - E E.................. '^rr< r->; — > l | | r XR~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~... ---..... 2,,.,.,.,.,.,.,.,.,,,,..............~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.............. <RGD - 2 DB ggg.-B., g,,,,,., - ' a..~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~...... w -. Rg-B.,g.,.,B B B - - B B.-' BB ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~~ ~~~~~~~~~~~~~~~~.................. ' igure 1.2 Filter type masks and protective ch~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~...................in.... (Courtesy of General Electric Co. )~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~..............

1-9.* Figure 1.3 High level radiation area at the Hanford Atomic Products Operation marked for protection of personnel (Courtesy of General Electric Co.) s~~~~~~~~~~~~~~~~~~~~~~~.....:. Fiur Pecls an bage wor to obai recrd of radi-.-:? ation exposure (Courtesy of General ElC..o............) I~~~~~~~~~~~~~~~~~~~.. 5. I.....; 0 l g | | | jj; w;;;;;;;," "l' S | l'i';i; f'VE~'t'S'S't'ii:0g~g'S:S'S~jg'S~t~f~tE'S:':;:'i;:'4,t~lt~i-.' Figure 1. 4 ig Pencils andito bre dt th -frdAoi Prdut Oeato ge wrne for o al ectord of praaton epouel (Courtesy of General Electric Co. )......................:r~~~~~~~~~~~~~~~............... B i-~iiiiil~............... Figure 1,4 Pencils and badges worn to obtain records of radi~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~Anile ation exposwre (Courtesy of General Electric Co.)~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~........

1-10 electromagnetic (X or gamma) radiation, and the approximate energy of the latter also may be evaluated. The developed film is read in a densitometer, and the amount of fogging is related to the amount of external radiation to which a person has been exposed. Sometimes a special film that will record neutron tracks in the emulsion is used to indicate exposure to neutrons. This film is usually placed inside the same badge that contains the betagamma film. (See Chapter 3 for additional information on film badges.) To obtain a better understanding of safety in working with radiation, the units of radiation as applied to health physics are discussed in the next section. 1.5 Radiological Units Various units of radiation dosage are employed in health physics, the most important of which are defined as follows. Roentgen The first unit of radiation was the roentgen, developed for X-ray measurements. It is a unit of energy absorption in air and defined for X- or gamma radiation only. The following equivalents should serve to clarify its significance: 1 roentgen = 1 ESU per c.c. of air (at STP) 20083 x 109 ion pairs per c.co of air - 6.95 x 104 Mev.per c.c. of air 5.40 x 107 Mev.per gram of air = 86 ergs per gram of air The roentgen is now the unit of exposure dose for X- and gamma radiation and is officially defined as "that quantity of X- or gamma radiation such that the associated corpuscular emission per 0.001293 gram of air produces in air one electrostatic unit quantity of electricity of either sign." (14) Rep (roentgen equivalent physical). The rep was the first attempt to extend the concept of absorbed dose to tissue. Several definitions of the rep have been proposed, but the most widely accepted one is that it is a dose in tissue of 93 ergs of ionizing radiation absorbed per gram of soft tissue. The rep is no longer in active use, having been replaced by the more general unit, the rad. Rad The rad is the present standard unit of absorbed energy. It is a measure of the dose of any ionizing radiation in terms of the energy absorbed per unit mass of the absorber. One rad is equal to 100 ergs of energy absorbed per gram of absorber. It should be noted that the rad may be used with any material such as tissue, metals, chemicals, etc. that have very different absorption characteristics.

Rem The rem is the dose of any ionizing radiation to body tissue in terms of its estimated biological effect equivalent to the dose in tissue of one rad of X-radiation having a linear energy transfer to water of 3.5 Kev per micron. The relation of the rem to other dose units depends upon the biological effect under consideration and upon the conditions of irradiation. For purposes of comparison, any of the following may be considered as equivalent to a dose of one rem: (1) A tissue dose resulting from 1 r due to X- or gamma radiation; (2) A tissue dose from 1 rad due to X-, gamma, or beta radiation; (3) A tissue dose from 0.1 rad due to neutrons or high energy protons; (4) A dose in tissue of 0.05 rad due to particles heavier than protons and with sufficient energy to reach the lens of the eye. However, it should be born in mind that any exact relationship between the roentgen, the rad, and the rem is very difficult to express for several reasons. (1) The biological damage is nearly always basically different for the several kinds of radiation. Therefore, the rem should never be used to express results of biological experiments. (2) The rep and rad units are defined as absorbed energy in any gram of matter whereas the roentgen is defined in terms of ion pairs produced per unit volume of air by the secondary electrons. (3) It is not known whether or not damage to tissue is proportional to the number of ion pairs produced or if the damage is proportional to the energy expended in the tissue, Figure 1.5 illustrates schematically the greater local tissue damage caused by "softer" radiation such as alpha particles, beta particles and soft X-rays. Dobsages besulting from neutrons are more conveniently measured and expressed in terms of the neutron flux than in rads or rems. For rough estimates one rem of neutron dosage may be assumed to be equivalent to the exposure of the body to 14 million neutrons per square centimeter, If the approximate energy distribution of the neutrons is known, a more precise determination of dosage is possible with the use of Table 1.5, Table 1.5 gives both the number of neutrons per square centimeter equivalent tio a dose of one rem and the average flux (neutrons) required to deliver 100 millirem in 40 hours as a function of neutron energy.

1-12 AlphO Partices Gamma Rays Beta Particles Hard X- Roys Soft X- Rays \ ~ \ Surface of tissue M/Teineters,.. Figure 1.5 Effect of different types of radiation on tissue (15) 100 - 90 l 80 g 60 z 50 40 0 100 300 500 700 900 L-D 50 of 4.5 roentgens (for man)

1-13 Table 1.5 Neutron Flux Dose Equivalents (16) Neutron Energy Number of Neutrons per cm.2 Average Flux to Deliver 100 Mev Equivalent to a Dose of 1 Rem Millire,ms in 40 Hours (neutrons/cm.2)X 106 (neutrons/cm.2 per sec.) Thermal 970 670 0.0001 720. 500 0.005 820 570 0.02 400 280 0.1 120 80 0.5 43 30 1.0 26 18 2.5 29 20 5.0 26 18 7.5 24 17 10.0 24 17 10 to 30 14 10 The foregoing definitions serve to identify quantities of radiation effects, either as ionization in air, energy in tissue, or.biological damage. It is further of interest to define a unit of quantity. of radioactive material. Curie One curie is a quantity of a radioactive nuclide in which the number of disintegrations per second is 3.700 x 1010. This unit is entirely independent of decay scheme or type of radiation emitted. The radiation level to be expected in the vicinity of a curie of radioactive material will depend only on the type and energy of radiation emitted in addition to external factors. Through the years it has been convenient to speak of a gram of radium, or the amount of radioactivity associated with a gram of radium, This amount of radioactivity has been defined as a curie, in recognition of the fundamental contributions of Marie Curie in the field of radioactivity. More and more precision has been applied to the experimental determination of the number of disintegrations per second from one gram of radium. Recent measurements vary from 3.4 to 3.7 x 1010, To eliminate this uncertainty and specify the unit of radioactivity, the valua g iven above! i;iS:;now:..in'ageneral use.

1-14 L.6 Biological effects of radiation The absorption of ionizing radiation by living tissue results in damage to individual living cells. This damage is followed by malfunction or death of the irradiated cells and may subsequently cause the death of the organism. However, no pain or discomfort is experienced by the victim at the time of the exposure until doses much greater than a lethal dose are received. Death does not follow immediately after whole body exposure to lethal doses of radiation. Destruction of vital cells, build-up of degenerative toxins, and lowered resistance to trauma and infection occur. Death results from secondary effects (19, 20). 1.7 Radiation syndrome The clinical manifestations produced by excessive exposure to ionizing radiation, often referred to as the "radiation syndrome", have four phases. The first phase, including the early symptons of nausea and vomiting, followed by general lassitude, is the "radiation sickness" of patients receiving intense radiation therapy. The second period may last from a few days after severe exposures to several weeks following lesser exposures and is a period of general wellbeing. This is followed by the third and crucial period in which the body reaction reaches a maximum. The patient loses appetite and weight and experiences general prostration, palpitation of the heart, bleeding of the gums, loss of hair, and severe diarrhea. In mild cases this phase may last only days, but in severe cases it may last weeks or become progressively worse until the patient succumbs. If the patient survives the third stage, a period of convalescence, which is the fourth stage, follows. In spite of the patient's apparently complete recovery from the radiation syndrome following massive sublethal radiation exposures, some permanent physiological impairment may be sustained. Extensive animal irradiation experiments, as well as follow-up data on the survivors in Hiroshima and. Nagaski, indicate two aspects of this long-term hazard. Morgan (1) listed five factors which determine radiation damage to man: (1) Total accumulation of exposure. (2) Area or volume of tissue irradiated. (3) Radio-sensitivity of body tissue involved. (4) Rate of irradiation. (5) Specific ionization of the radiation.

108 ' i 0 / x x z0 -CO IOCM IO /I __/ Z -4...G _yor isotopic source o- fast (10 to 30 Mev) neutrons. Prepared by

The total X- or gamma ray exposure required to kill 50 per cent of a group of humans has been estimated to be between 400 and 500 roentgens. This estimate is based on animal experiments and a few large human exposures (19). The biological effects of various doses of total body exposure and local exposure are summarized in Table 1.6 (20). For additional information on the biological effects of radiation exposure see references 21 to 89. Table 1.6 Radiation Dosimetry and Biological Response (18) Dosimetry Dose r Dose Rate Exposure* Biological Response 0.3 Weekly T Probably none (permissible dose) 1 Daily (for years) T Leukopenia 1.5 Weekly L Probably none (permissible dose for hands and fingers) 25 Single dose L Chromosome break in tumor cells (tissue culture) 50-100 In accumulated L Gene mutations to double spontanesmall dose ous rate per generation 60 Single dose L Depression of phosphate activity 200 Single dose T Nausea 300 Single dose L Erythema dose for 100 kv (small field) 300-500 Single dose T LD50 for man 300-600 Single dose L (ovaries) Sterilization in female 400 Single dose L Reversible epilation 400-500 10-50 r/day T Clinical recovery 500 Single dose L Erythema dose for 200 kv (small field) 600o-oo Single dose L (testes) Sterilization in male 600-900 300 r/day or L Radiation cataract small doses 1000 Single dose L Erythema dose for radium 1000-1500 200-300 r/day L Epiphyseal retardatian 1000-2500 200-300 r/day L Response of markedly radiosensitive cancer 1500 200-300 r/day L (ovaries) Castration in female 1500-2000 200-300 r/day L Cessation of salivary glandular functions, 1800-2000 200-300 r/day L (stomach) Archlorhydria 2000 Single dose L Erythema dose for 2 Mev 2000-3000 200-300 r/day L (kidney) Radiation nephritis

Table 1.6 conto 2500-6000 200-300 r/day L Response of moderately radiosensitive cancer 2700-3000 Single dose L Moist desquamation, but healing of skin; 100-kv radiation (small field) 3600-5000 200-300 r/day L Limits of skin (single portal, 200 kv, 5 x 5-cm field) 4000-5000 200-300 r/day L Moist desquamation, but healing of skin (single portal 2000 to 3000-kv radiation, 10 x 10-cm field) 50,000 10-100 r/day L Carcinogenic * T = total body; L = local. 1.8 Effect on cancer incidence One of the most feared aspects of radiation is the possible increase in cancer incidence. Several cases of skin cancer developed following radiation injury to the skin of early X-ray technicians. Increased rates of cancer incidence in animals exposed to ionizing radiation have been reported by many laboratories. Lawrence and Hamilton have reviewed the field and report a summary of carcinogenic effects of local and whole body irradiation as well as of internally administered radioelementso (21) Equally important in this regard, but less obvious, is the observed increased incidence of leukemia following radiation exposure in experimental animals. The death rate from leukemia among physicians has also been found to be twice that of the total adult male population (22), and the death rate among radiologists to be ten times that found among physicians in general. These two figures are commensurate with the portion of physicians working in radiology (23). Although these data are not direct proof that leukemia in humans is induced by radiation, the existence of an increased incidence, which is well above statistical deviations, among persons exposed to ionizing radiation is clearly indicated. References 24-29 give additional information on carcinogenesis and leukogenesiso 1.9 Effect on life span Another aspect of long-term physiological hazard associated. with radiation exposure has to do with a possible shortening of the life span. Because the plot of the incidence of death against radiation dose -is an S-curve as shown in Figure 1.6 (for man), the 50 per cent point is well defined, but lower and higher mortality rates are not. For comparison of experiments, therefore, the 50 per cent fatality point is generally usedo

1-17 When a particular strain of mice were exposed to 400 r, 50 per cent of the irradiated animals died within 30 days (30). Thus, 400 would be called the L.D.-50/30). With further fractionation, the number of deaths directly resulting from the same total exposure were reduced to zero. However, the life spans of these irradiated animals were found to be significantly less than those of nonirradiated animals. The percentage reduction in survival time was found to be proportional to the size of the daily exposure in the range from 5 to 25 r per day. Yockey (31) suggests that this induced aging might serve as a well-defined measure of radiation damage. 1.10 Radiosensitivity of different tissues The variation in the radiosensitivity of body tissue may influence the extent of radiation damage. As early as 1913 the concept of tissue sensitivity depending on the rate of metabolism or rate of cell division was advanced (32). The available data on radiation damage support such a concept. Numbered among the more radiosensitive tissues of the body are the lymphocytes, gonads, bone marrow and the gastrointestinal tract, while nerve, muscle and heart tissues are less radiosensitive. A number of interesting experiments illustrative of this principle of radiosensitivity appear in the literature. Jacobson (33) gave the following data (Table 1,7) to illustrate the increased survival rate of rats when the spleen was shielded from radiation damage.. References 34-43 give additional information on the effects of radiation on special tissues. Table 107 The Sensitivity of Spleen in Albino Rats to Radiation Damage (33) Dose Survival of Survival of Nonshielded Controls Spleen Shielded 600 r 50.0% 100% 1025 1.2 86. 0 1100 0 55.0o 1200 0 14.3 1300 0 3.0

1-18 1.11 Effect of rate of irradiation Morgan (13) gives as a fourth factor of radiation damage the rate of dose delivery. Again one can illustrate this phenomenon with techniques employed by radiologistso The radiologist "fractionates" his X-ray therapy doses to prevent overwhelming the body's recuperative powers and also because certain tissues with high reproductive capacity are more affected by repeated doses. Kaplan (44) has illustrated the effect of dose fractionation on the 30-day L.D.-50 in mice. He found the L D. 50 for a single dose to be 510 roentgens, for two equal daily fractions the L.oD -50 was 585 r, for four equal daily doses it was 720 r and for eight it increased to 850 roentgens. He then tried equal fractionated doses four days apart and found the L.D.-50 increased even more. 1.12 -The influence of the nature of the radiation A fifth factor influencing radiation damage to man is that of specific ionization of the radiation~ It is known, for example, that low-energy Xradiation is more effective than high-energy X- or gamma radiation in producing local damage in tissue (see Figure 1.5')o Although -the characteristic cellular destruction caused by ionizing radiation is basically the same for all known types of ionizing radiation, the severity and degree of localization of the injury vary according to the rate of absorption of the type of radiation in question. External alpha radiation, for instance, is totally absorbed in the horny surface layer of dead epithelial tissue. As a result, significant injury to living tissue from external alpha radiation is extremely unlikely.~ Beta radiation, on the other hand, can penetrate several millimeters of tissue. Severe blistering and burns, which heal very slowly, result from local overexposure to external beta radiationa. Gross total body damage from external beta radiation occurs only if the dose has been very largeo The tissue damage resulting from an X-ray or gamma-ray exposure arises from the production of secondary electrons. The ionization density along the path of the secondary electron is a function of the electron energy, which depends on the incident photon energy. The effects of gamma radiation, because of its highly penetrating nature, are produced throughout the body rather than locally. Gross metabolic effects, such as reduction in white blood-cell levels, result from severe overexposure. The neutron, though not an ionizing particle, produces recoil protons which give rise to dense ionization tracks in tissueo There is evidence which seems to indicate that certain organs and tissues in the body are especially sensitive to neutron exposures (45-47)~

1-19 Exposure to neutrons introduces a different type of effect in addition to the dense ionization tracks which occur at points of interaction between the neutron and atomic nuclei. Under certain conditions a neutron is. absorbed by the nucleus, resulting in the production of an "excited" or radioactive nucleuso This phenomenon presents a two-fold hazardo First, additional radiation exposure results from this induced radioactivity. Second, the nucleus in question is transmuted to an element higher in the periodic table. If the atom is part of a molecule which plays an important role in cellular metabolism, it may be anticipated that cellular function will be impaired. Table 198 summarizes the health-physics characteristics of various types of nuclear radiationso References 48-73 give additional information on whole body irradiation. Table 1.8 Health-Physics Characteristics of Various Nuclear Radiations Alpha Beta Gamma Neutrons Tissue Surface A Few Many Many Penetration of Skin Millimeters Centimeters Centimeters External None Severe Gross Radi- Gross RadiInjury Burn ation Effects ation Effects External None 1/4" Lead as High CrossShielding Lucite Required section Material Required as Required Internal Cell Cell Gross Gross Injury - Destruc- Destruc- Radiation Radiation tion and tion and' Effects Effects Tumor' Tumor ProProduction duction

1-20 ~.13 Internal hazard When a radioactive element enters the body it is distributed throughout the body, metabolized and utilized in the same manner as it would be if it were non-radioactive (74). When these processes lead to incorporation into the body structure by- de-position in the bones or other vital organs, localized internal radiation damage may ensue (75). As in the case of external radiation injury, no immediate pain or discomfort is experienced by the victim, sometimes not for several yearso This time lag between the exposure and the first clinically observable symptoms renders the tasks of diagnosing the initial cause of the injury and evaluating potential damage from an exposure exceedingly difficult. Radioactive material, when injested, presents a greater hazard than when the body is exposed to external sources. The reasons are as follows: (1) Radioactive nuclides within the body are in intimate contact with the surrounding tissue. Thus, alpha and beta radiations can dissipate all of the energy in a small volume of tissue. (2) It is extremely difficult to determine the amount of material in the body precisely. (3) The body is irradiated continuously until the substance is eliminated. (for example: 10 juc Ra226 inJested would result in an average dose rate to the bone of 56 rem/wh or about 151,000 rem assuming a 50 year life after injestion) (4) It is impossible at the present time to satisfactorily remove injested radioactive material by artificial means. Therefore, it is very important that persons working with radioactive material avoid any possibility of injestion of radioactive material. Significantly the amount of radium in the preceeding example would weigh 0.00001 grams. Not all radioactive materials however, are as hazardous to handle as radium since some nuclides have an extremely short physical half life and others have a short biological half life in addition to other factors. Even so, a person should always avoid needless injestioneof any radioactive material because any amount of exposure to radiation may do some harm. ~.14 Possible effect of radiation on subsequent generations In a sexually reproducing species, such as man, the transmission of all

1-21 hereditary characteristics from generation to generation is maintained through one cell from each parent. These two cells fuse and develop to form the new individual. Radiation damage to these sex cells, or to the cells responsible for their production, may produce abnormalities in the new individual which can be transmitted to the next generation. Such changes are called mutations. As a mutation is a permenent hereditary change, and may be caused by a single ionizing event, the total number produced in an individual will depend on the total dose received. Russell (83) has indicated, however, that there is some dependence on the dose rate, the mutation frequency being lower at the lower dose rates. The mechanism of inheritance is one of the most radiosensitive biological functions. The "Digest of Findings and Recommendations" of the National Academy of Sciences states that any quantity of radiation, however small, may cause mutations (84). The most recent report by this group states that genetic effects per unit of dose rate may be less than previously estimated. However, there is still considerable question as to what extent mutations are increased by chronic exposure to radiation. All individuals are exposed to background radiation, which causes an unavoidable background mutation rate. The number of mutations present in an individual is cumulative and builds up as the total radiation exposure increases from the time of conception of the individual until conception of his last child. Mutant genes can survive until the inheritance line in which they are carried dies out. If the mutation involves a serious loss of function, the line may cease after the first filial generation. If the mutation effect is slight, it may be carried through hundreds of generations before elimination. Radiation-induced mutations are not new genetic changes, but constitute an increased rate of occurrence of normally occurring mutations. Favorable mutations have been absorbed into the population and unfavorable mutations eliminated through natural selection. Most new mutations are likely to be deleterious to the individual born with them. About 50 per cent of all children are born to parents under 30 years of age and about 90 per cent to parents under 40 years of age. It has been estimated that a typical individual at the age of 30 would have received a cumulative radiation dose of 4.3 rem from background; 3 rem from X-radiation and fluoroscopy; and.02 and 0.5 rem from fall-out from atomic weapon tests.

1-22 1.15 Shortening of life span There is increasing evidence that radiation in large dose may lead to shortening of the life span of an individual aside from the result of damage to a specific tissue such as development of skin cancer, leukemia, etc. It is also thought by many pathologists that there is no "threshold. effect" for this shortening of the life span. Thus, presumably any dose of radiation, however small, may shorten expected life by some small fraction. It is believed however, that the levels of radiation as recommended by the NCRP and the ICRP are such that any shortening of normal expected life span will be insignificant. One may well expect greater effect from air polution or any number of other environmental factors. 1.16 Significance of radiation effects The immediate reaction of the popular press in 1957 to the report "The Biological Efrects of Atomic Radiation" by the National Academy of Sciences was that of apprehension (100). The reaction of persons familiar with the situation, on the other hand, was favorable to the report and definitely not apprehensive (101). It is felt that the report notably documents and assembles already existing information about radiation effects, and that sufficient time remains before the nuclear power industry reaches its anticipated large size to evaluate, verify and prepare intelligently for the degree of radiation hazard to be expected. In 1958 Russell et al. (83) reviewed the effect of dose rate on mutation frequency. Additional information on radiation hazards are in References 84 to 99. The nuclear industry has been so conservative since its inception that the average exposure of the personnel working in -,he present atomic energy industry has been much less than the internationally accepted maximum permissible levels iinuse since the discovery of fission. Thus, downward revisions of these levels should not cause undue concern. Perhaps the most important recommendations of the report were that total exposure to medical X-rays be reduced by all means possible to the minimum consistant with medical necessity, and that records of all important medical exposures be maintained in addition to records already maintained for exposure to nuclear radiations. 1.17 Maximum permissible exposure levels Maximum permissible radiation exposure levels are based on an intensive study of the biological data now available, although according to one theory, any dose of ionizing radiation may cause some somatic or genetic damage. The levels set are considered to entail a risk no greater than is

1-23 presently accepted in other industries. In the early days of work with radiation, when the possibility of injury from radiation exposure was becoming evident, any safety measures that existed were self-imposed. One prominent worker in the field monitored his exposures by clipping a piece of photographic film to his lapel with a paper clip. If a certain radiological procedure resulted in the appearance of a photographic image of the paper clip, he would revise the procedure to reduce future exposures. His early "tolerance"- dose was, therefore, any amount of radiation which would not produce a photographic image of a paper clipo As industrial, medical, and laboratory uses of radiation expanded, a growing recognition of the radiation hazard occurred. During the 1920's and 1930's local standards of radiation tolerance were slowly established. These local standards provided valuable guidance to medical and health-physics people when.they faced the tremendous radiation problems associated with the Manhattan Project. As radiation intensity decreased with the square of the distance from a radiation source and with the amount of shielding material-between the source and the observer, sufficient distance or shielding thickness can reduce the radiation from a particular source to any desired level. In the practical case, space and monetary limitations impose some lower limit on. the minimum attainable radiation level. It is impossible, of course, to reduce radiation levels below the level due to cosmic rays and the presence of natural radioactivity in everyday materials. It was important, therefore, to define a rate of radiation exposure which could be tolerated by an individual for any given period or for a lifetime with no observable ill effects. Such a "tolerable" dose has come to be called a "tolerance doseo" The National Committee on Radiation Protection adopted the limit for external radiation of 0.1 r for an 8-hour day in 1936. The permissible radiation exposures in Britain and Canada during the same interval were 0.2 and 0.05 r per day, respectively. These values.have since been loweredseveral times. Table 1.9 shows how the "tolerance dose" for individuals who work with radium or X rays, or in the atomic energy industry has been modified since 19Q2.

1-24 Table 1.9 Historical Development of- "Tolerance Dose" (102) Author Date Calculated r/day Rollins.1902 10 Mutscheller.,1925. 0.2 Sievert 1925 0.2 Solomon 1926 2.0 Dutch Board 1927 0.04 Bayclay and Cox 1928 0.17 Kaye 1928 0.12 Failla ( rays) 1932 (2) 0.1 Stenstrom 1932 o.-16 ICRP (3) 1934 0.2 NCRP (4) 1936 0.1 NCRP 1949 - (5) 0~.075 ICRP 1950 - (5) 0.075 ICRP 1958 0.-025 NCRP 1959 0.025 (1) Photographic;(2) Established by ray erythema? 1,800 r;other calculations or erythema - 600 r;(3) International Commission on Radiological Protection, then known as International X-ray and Radium Protection Commission; (4) National Committee on Radiation Protection-and Measurements, then known as Advisory Committee: on X-ray and Radium Protection; (5) Based on a weekly irradiation of 0.3 r in air. The fact that the "tolerance" level has been steadily lowered gives no assurance that a lower limit has finally been reached Future revisions could be even lower. The concept of "maximum permissible" dose rather than a so called "tolerance dose" was first applied in handbook 59 of the National Bureaui of Standards(. t:This new terminology was used instead of the older term because "tolerance" indicates an assumption that if the dose iss lower than a certain value-no injury results. Since recent investigations show there is no "threshold" dose for gene mutations or somatic damage, the term "maximum permissible" dose has been accepted implying there is some finite risk involved yet so small that the risk will be readily acceptable. The International Commission on Radiological Protection (ICRP) and the National Committee on Radiation Protection (NCRP) in 1958 published virtually identical recommendations regarding maximum permissible limits for external and internal radiation exposures. ( L0O3-1Q6)

The following is a reproduction of part of the addendum to handbook 59 and is a summary of the recommendations as regards external exposure. (i06) 1.18 Accumulatedl dose (radiation workers)(106) A. External exposure to critical organs Whole body, head and trunk, active blood-forming organs, or gonads: The maximum permissible dose (MPD), to the most critical organs, accumulated at any age, shall not exceed 5 rems multiplied by the number of years beyond age 18, and the (dose in any 13 consecutive weeks shall not exceed 3 rems * Thus the accumulated MPD - (N-18)X 5 rems, where N is the age in years and is greater than 180 Comment: This applies to radiation of sufficient penetrating power to affect a significant fraction of the critical tissue. B. External exposure to other organs Skin of whole body: MPD _ 10 (N-18) rems, and the dose in any 13 consecutive weeks shall not exceed 6 rems.**Comment: This rule applies to radiation of low penetrating power. See figure 2, H59. Lens of the eyes: The dose to the lens of the eyes shall be limited by the dose to the head and trunk (A, above). Hands and forearms, feet, and ankles: MPD - 75 rems/year, and the dose in any 13 consecutive weeks shall not exceed 25 remso*** C. Internal exposures The permissible levels from internal emitters will be consistent as far as possible with the age-proration principles above. Control of the internal dose will be achieved by limiting the body burden of radioisotopes. This will generally be accomplished by control of the average concentration of radioactive materials in the air, water, or food taken into the body.. The quarterly limitation of 3 rems in 13 weeks is basically the same as in H59 except that it is no longer related to the old weekly dose limit. The yearly limitation is 12 rems instead of the 15 rems as given in the NCRP preliminary recommendations of January 8, 1957. ** This is similar to the 1954 (H59) recommendations in that the permissible skin dose is double the whole-body dose. H59 made no statement regarding a 13-week limitation. ~X *This is basically the same as the 1954 (:H59) recommendations except for the 13-week limitation.

1-26 Since it would be impractical to set different MPC values for air, water, and food for radiation workers as a function of age, the MPC values are selected in such a manner that they conform to the above-stated limits when applied to the most restrictive case, vizo, they are set to be applicable to radiation workers of age 18. Thus, the values are conservative and are applicable to radiation workers of any age (assuming there is no occupational exposure to radiation permitted at age less than 18). The maximum permissible average concentrations of radionuclides in:air and water are determined. from biological data whenever such dataare available, or are calculated on the basis of an averaged annual doseof 15 rems for most individual organs of the bod.y,* 30 rems when the critical organ is the thyroid or skin, and 5 rems when the gonads or the whole body is the critical organ. For bone seekers the maximum permissible limit is based on the distribution of the deposit, the RBE, and a comparison of the energy release in the bone with the energy release delivered by a manimum permissible body burden of 0.1ol ig Ra226 plus daughters. 1.19 Emergency dose (radiation workers) (106) An accidental or emergency dose of 25 rems to the whole body or a major portion there-of, occurring only once in the lifetime of the person, need not be included in the determination of the radiation exposure status of that person (see p. 69, H59).** 1.20 Medical dose (radiation workers) (106) Radiation exposures resulting from necessary medical and dental procedures need not be included in the determination of the radiation exposure status of the person concerned..**' 1.21 Dose to persons outside of controlled areas (106) The radiation or radioactive material outside a controlled area, attributable to normal operations within the controlled area, shall be such that it is improbable that any individual will Deceive a dose of mote than' 0S rem in any'l year from external radiation. The maximum permissible average body burden of radionuclides in persons outside of the controlled area and attributabie to the operations withein the controlled area shall not exceed one-tenth of that for radiation workers.o**** This is.basically the same as the 1953 (H59) recommendations~ (117) ** This is the same as the 1954 (H59) recommendations (117) *** This is basically the same as the recommendations of January 8, 1957 (105)

1-27 This will normally entail control of the average concentrations in air or water at the point of intake, or rate of intake to the body in foodstuffs, to levels not exceeding one-tenth of the maximum permissible concentrations allowed in air, water, and foodstuffs for occupational exposure. The body burden and concentrations of radionuclides may be averaged over periods up to 1 year. The maximum permissible dose and the maximum permissible concentrations of radionuclides as recommended above are primarily for the purpose of keeping the average dose to the whole population as low as reasonably possible, and not because of specific injury to the:.individual. Comment: Occupancy-factor guides will be needed by several of the subcommittees. It will be important that these do not differ markedly between different handbooks. The Executive Committee will endeavor to establish a set of uniform occupancy-factor guideso 1.22 Operational and administrative guides The maximum dose of 12 rems in any 1 year as governed by the 13 week limitation, should be allowed only when adequate past and current exposure records exist. The allowance of a dose of 12 rems in any 1 year should not be encouraged as a part of routine operations, it should be regarded as an allowable but unusual condition. The records of previous exposures must show that the addition of such a dose will not cause the individual to exceed his age-prorated allowance. The full 3-rem dose should not be allowed to be taken within a short time interval under routine or ordinary circumstances (however, see paragraph 2 on Emergency Dose above.) Desirably, it should be distributed in time as uniformly as possible and in any case the dose should not be greater than 3 rems in any 13 consecutive weeks. When the individual is:.not personally monitored and/or personal exposure records are not maintained, the exposure of 12 rems in a year should not be allowed; the yearly allowance under these circumstances should be 5 rems, provided area surveys indicate an adequate margin of safety. When any person accepts employment in radiation work, it shall be assumed that he has received his age-prorated dose up to that time unless (1) satisfactory records from prior radiation employment show the contrary, or (2) it can be satisfactorily demonstrated that he has not been employed in radiation work. This is not to imply that such an individual should be expected to routinely accept exposures at radiation levels approaching the yearly maximum of 12 rems up to the timel..he reaches his age-prorated limit. Application of these principles will serve to minimize abuse.

1-28 The new MPD standards stated above are not intended to be applied retroactively to individuals exposed under previously accepted standards. It is implicit in the establishment of the basic protection rules that at present it is neither possible nor prudent to administer a suitably safe radiation protection plan on the basis of yearly monitoring onlyo It is also implicit that at the low permissible dose levels now being recommended, there is fairly wide latitude in the rate of delivery of this dose to an individual so long as the dose remains within the age-prorated limits specified above0 In spite of a lack of clear evidence of harm due to irradiation at dose rates in excess of some specified level, it is prudent to set some reasonable upper limit to the rate at which an occupational exposure may be delivered0 Therefore, it has been agreed that the dose to a radiation worker should not exceed 3 rems in any 13 consecutive weekso The latitude that may appropriately be applied in 'the operational and administrative control of occupational exposure will be dictated by two major factors (a) the type of risk involved and the likelihood of the occurrence of over-exposures and (b) the monitoring methods, equipment, and the dose redording procedures available to the radiation userso Where the hazards are minimal and hot likely to change from day to day or where there are auxiliary controls to insure that the 13-week limitation will not be exceeded, the integration may be carried out over periods up to 3 months~ Where the hazards are significant and where the exposure experience indicates unpredictability as to exposure levels, the doses should be determined more frequently, such as weekly, daily, hourly, or oftener, as may be required to limit the exposure to permissible valueso For the vast majority of installations (medical and industrial), operation is more or less routine and reasonably predictable and it may be expected that their monitoring procedures will be minimalo For such installations the protection design should be adequate to insure that over-exposures will not occur —otherwise frequent sampling tests should be specified0 Where film badges are used for monitoring, it is preferable that they be worn for 4 weeks or longer, since otherwise the inaccuracy of the readings may unduly prejudice the radiation history of the individual. Where operations are not routine or are subject to unpredictable variations that may be-hazardous, self-reading pocket dosimeters, pocket chambers, or other such devices should also be worn and should be read daily or more often as circumstances dictate0 Except for planning, convenience of calculation, design, or administrative guides, the NCRP will discontinue the use of a weekly MPDor MPC.*: The Committee has deliberately omitted the discussion of future exposure forfeiture for exposures exceeding the MPD on the grounds that any such statements might lend encouragement to the unnecessary use of forfeiture provisions. *This represents a minor change from the NCRP recommendations of January 8, 1957, but no change in the basic MPDo

1-29 The maximum permissible body burdens and maximum permissible concentrations of radionuclides as specified by the Rules and Regulations from Title 10, Part 20 of the Federal Register are given in Table 1.10 (107). Table 1.10 Maximum Permissible Concentrations for Radioisotopes in Air and.Water Above Natural Background (107) Table I Table II Element (atomic number) Isotope 1 Column 1 Column 2 Column 1 Column 2 Air Water Air Water (ac/ml) (Pc/ml) (ac/ml) (pa/ml) Actinium (89) ------------------- Ac 227 8 2X10-1 6X10-6 8X10-1 2X10-4 I 3Xl -1 9X 10-8 9X 10-I 3X10-'.c 228 8X10-' 3X10-' 3X10-o 9X1! -S I 2X10-6 3X10-8 6X 10-" 9X o10 -Americium (95)-.................Am 241 8 6Xle-" 1X10e- 2X10-le 4X1i-4 I 1 X 10-1 8X 10- 4X 10s 2Xl1 -Am 243 8 6X 1e0 -e 1X10" 2X 1 -18 4X1O4 I 1 X10-1 8X10' 4X1e0- 3Xl1Oe Antimony (51) -S............... 8b 122 8 2X10e- 8X 10-4 6X1O 3XIOI 1X 1O-7 8X 10 X 10x- 3X10-' Sb 124 8 2X10-' 7X10-' 5X1e-" 2aXie-' I 2X 10- 7X 1-4 7X 17X- e- 2X 10 -8b 125 8 5X1017 3X10-' 2Xe10- 1Xi0-' I 3X10- 3X1a x ' 9X 10x-1 eX1-4 Argon (18) ------------------------ A 37 Sub' 6XIO - - 1X10-4 A '41 Sub 2X10-4 4X10 --- Arsenic (33)..-............... As 73 a 2X10-4 1Xi'2 7X1o-, 5X1iI 4X 10-7 1X10-x- 1X0- 5X1O-4 As 74 8 3X10-' 2X10-8 1Xi10- 5X 1 lO I 1X10-7 2X 10- 4iX10- 5X10-' As 76 8 1X1O-7 6X 1O' 4X 10- 2XI10 -I 1X 10- 6X1eO- 3X10i- 2X10-6 As 77 8 5X1e-7 2X1e-3 2X10-' 8X l(I 4X1-'7 2X1eO- iXIe0' 8X10"Astatine (85) ------------------- At 211 S 7X1I0- 5X1e-1 2X1I0-o 2X10-4 I 3X10-' 2X10-3 1X 10- 7Xe10 -Barium (56) -------------------- Ba 131 8 X 10- 5X1e0- 4XI — 2XO1 -I 4Xe10- 5X10- 1Xe10-' 2X-104 Ba 140 8 1X1e-7 8X1O-' 4X10-0 3X 10 -I 4X10-' 7X1(0- 1XlO- 2X1eOBerkellum (97) --------------- - Bk 249 9 9X10el- 2X10 — 3XIO-ll X1)e-'4 I 1X10-7 2X10-s 4X1O- 6X10-4 Beryllium (4).. ----—. Be 7 8 6X10- 5X10-t 2X1e-7 2X1iI lX 10' 5X10-' 4X10- 2X1e0 -Bismuth (83) - -------- Bi 206 8 2XIO-l' IX10-' 6X010- 4XIe0-' I 1 X 10- 1Xl1- 5X-10e 4X1e0 -Bi 207 8 2Xle-' 2X107 2 X 10- 6X 10xlI 1X l10 2X10-' 5X10-"~ 6X10-' Bi 210 8 6X10-' 1XIe- 2X10-1o 4X10-' I 6X10e- 1Xle — 2X1IOel 4X1e-' Bi 212 s x 10-7 1X10- 3Xle-i 4X10-4 I 2X10-7 1XX10- 7X1e-' 4X10-4 Bronmine (35) --- —-------------- Br 82 8 1X10- 8X10- 4X1- 3X10 -I 2X10- 1 X 10- 6X10 — 4X1e-5 Cadmium (48)- ---------------- Cd 109 8 5XI10- 5X10-t 2X10-' 2X10-4 I 7X1e-' 5XIO-S 3Xl1O- 2X1O-4 Cd 115m 8 4X1 0- 7X1e0- 1X 1-' 3X 10-s I 4X10-' 7X10-I lX 10- 3XlI-' Cd 115 8 2XlO-7 X1ie-* 8X10- 3X le-' I 2X10-7 IX10- 6X 10-o 4X1e0 -Calcium (20)-.._ -- -----—. Ca 45 8 3XI0- 3X10-4 1XIOe- 9X 10 -I 1X10-7 5X1e-1 4X10-4 2Xl104 Ca 47 8 2X10-t IXi1O- 6X1e-o 5Xa1 — I 2X1e7 1X10-x 6X 10e- 3X 1-,4 Californium (98) -........ Cf 249 8 2X0ltls lXl0- 5XslO-e 4X10e I 1X10-l 7X1e-' 3X10-Is 2X10-' Cf 260 8 SXlt- X10-ie' 2X10-Il 1X1e-' I 1X 10-lo 7X 10' 3X 10-Is X 10 -Cf 252 5 2X10-l 7X10-' 7X1Il- a2X10' I 1X 1-"1 7X10e 4X1e-t 2X1-'4 Carbon (6)-.......... C 14 8 4X10- 2Xl-S lX10e- 8X10' (COs) Sub SX1e- 1 X i -.., Cerium (58)..-....... Ce 141 8 4X10- 3X1s' 2X10-x 9X10e I 2X10-? 3X10-' 5X 10-s 9X10 -Ce 143 S 3X10- 1Xi 10- 9X10- 4X10-' I 2X10- 1iXe10- 7X10' 4X10-s Ce 144 S 1XI-' 3X10- 3X0I1e- IX10-' I 6X 10-' 3X10-' 2X10-l~ Xe10-s See footnotes at end of table.

1-30 Table 1.10 Contd. Table I 'rable II Element (atomic number) Isotope 1 Column I Column 2 Coluiin I Column 2 Air Water Air Water (c/Mnl I) (clm l) (Oe/m I) Ouc/ml) Cesium (55) ------------------—. Cs 131 S 1X10- 7X102 4XI0-7 2X10-3 I 3X10-6 3X10- l X 10-7 9X10"' Cs 134m S 4XIO-5 2> '0-i 1X10-0 6X1I0 -I f6X10-'6 3X 10-2 2XI0-7 1X 10 -Cs 134 S 4X10-8 3X10-4 1Xl(- XI0-6.I 1X10-8 1X10-3 4X10-1' 4X0-$s Cs 135 S 5XI0 3X10-3 2X10' -IX10-4 I D9XlO-s 7X 10-3 3X O-9 2X10-4 Cs 136 S 4X10)- 2X1()-3 1XIO-8 9Xl10 -- I 2X 10- 2X-3 6X 10-' 6X 10 -Cs 137 S 6X10-. 4X0- 2X10 2X10-9 2X I 1X lO 1 X 10 -3s 1XI l 5X 10-t 4X10-$ Chlorine (17).-.................. C136 S 4X10-7 2X10-3 1Xl O- 8X 10 — I 2X10-s 2X10-3 8X10-o0 6X 10-3 Cl 38 S 3X 10-6s X 1(-2 PX 10-8 4X 10 -I 2X10 I 1X 1(2 7TX10-l 4X 10-" Chromium (24) Cr51 S 1X10- 5X10-2 4X10-7.. 2X10-a. I - 2 X 10-a 5XIO-2 8X 1O- 2X 10 -Cobalt (27) -------------- ------ Co 57 S 3X10-6 2X10-2 lXI0-7 5X10-:I 2X10-17 1 X 0-2 Xi 10- 4X 10 -Co 58m S 2X I0-l 8X10-2 6X 10-, 3X10-t I 9X 10-6 6X 1O2 3X 10-7 2X 10-3 Co 5 S SX 10-7 4X10-3 3X10- IX10-4 I I 5X XI-h8e 3X10-3 2X 10-0 9X10 -Co 60 S 3X 10-7 1XI0-3 I X 10-8 5X105 -I X 1 (" 1X 0 I0'-3 3X 10-1O 3X 10 -Copper (29) ------------------- Cu 64 S 2X10-8. I X10-2 7X10-s 3X10-' I X 10X-6 iX 10-3 4X 10- 2X10-4 Curium (96).-.................... Cm 242 S 1X10-o 7X10-4 4X10-12 2X10-5 r I' 2X 10-1 7X 10- 6X 10- 3X105 -Cm 243 S 6iX IO- I Xt0-4 2X 10-t3.X. X0 -I l1X1O-1o0 7 X 10- 3X10-12 2X10-6 Cm 244a S 9X10-I- 2XI(0- 3X10-13 7X10 -I l X10-1o 8X 10-4 3X10-"' 3X10-$ Cm 245. S 5X1-l 1X10-4 2X10-13 4XI10 --I 1XI10-o 8Xl(0-( 4 X 10- 3X10-A Cm 246 S 5X 10-I2 1X 10-' 2X10-13 4X10' -I 1XIO-'1 8X 10- 4X10-12 3X10-' Dysprosium (6f) ------------ Dy 165 S 3X10-I II X0-2 9X10-e 4X10-4 I 2X10-s 1X10-2 7X 10- 4X10-4 Dy 166 S 2Xl0- 1X10-3 8X10-'o 4Xl0o-$ I 2X 10-'! X10-3 7X 10- 4X 10( Erbium (68) --------------------- Er 169 S 6X10-? 3X10-3 2X10-S 9X10-s I 4X 10- 3X 10- 1X Ii-6 9X 10 -Er 171 S 7XI1-0 3X10-3 2X10-8 1X 10 -I ( 6X10-7 3X10-3 2X10-s I X 1 - Europium (63) -. --- ——. --- --- Eu 152 S 4X10-7 2X10-3 lX 10-s. 6X10-' (T/2=9.2 hrs) I 3X10-t 2X 10- 1XI1- 6X10-S Eu 152 S 1XIl-s 2X10-' 4X10-. 8X10-'$ (T/2-13 yrs) I 2X0-s 2X 10- 6X10-1S ' 8X10-3 Eu 154 S 4X10- - 6X10-4 1XI10- 2X10 -I 7X 10- 6X 10- "2X 10-0" 2X 10-$ Eu 155 9 X10-8 6X10-" 3X 10- 2X10-4. I 7X10-s 6X 10-3 3X10-' 2X10-4 Fluorine (9) -. --- —--- --- - F 18 S 5X10-. 2X10- 2X10-? 8X10-4 I 3X10-' 1 X 10- 9X10-s 5X 10 -Gadolinium (64)..-G —........... Gd 153 S 2X10-' OX X10- 8X10-9 2X10-4: I 9X10-'8 6X10'" 3X10-' 2XIO" Gd 159 S 5X1(0-. 2X10-s 2X10 —. -. i8X10-s I. 4X10-t' 2XI- 10 X1O- 8X 10 -Gallium (31) --------------------- Ga 72 S 2X10-t 1X0-3 8X10 - 4X10-5 I 2X10-7 X 10- 6X 10'o 4X10-3 Germanium (32) --— G --- —----- Ge 71 S 1X 10-{ 5X 10- 4X 10-t 2X 10 -Il: 6X10-6 5X1(-2 2X10-T 2X10-3 Oold (79).-............... Au 196 S 1X10-': 5X10-3 4X10-8 2X10-' I 6X110-) 4X10'- 2X10-s 8 1 X1 -Au 198 S 3X10-? 2X10- 1X10-: 5X10-' I 2X10-t7 IX1s 8X10-' S X10-s

1-31 Table 1.10 Contd. Table I Table II Element (atomic number) Isotope ' Column 1 Column 2 Column I. Column 2 Air Water Air Water (jc/ml) (Ac/ml) (Gc/ml) (c/ml) Gold (79) ----—._ --- —- --- Au 199 S 1X106 5X10-3 4X10-I8 2X10-4 I 8X10-7 4X10-3 3X10-e 2X10-4 Hafnium (72) ----------- --- Hf 181 S 4X10-8 2X10-3 1X10-O 7X10-5 I 7X10-8 2X10-3 3X10 — 7X10-A Holmium (67) ----------------- Ho 166 S 2X10-7 9X10-4 7X10-9 3X10-s I 2X10-7 9X10-t 6X10-' 3X10-s Hydrogen (1) ------------------ H3 S 5X10- 1Xi10-' 2X10-7 3XIO-s Sub 2X10- -- 4X10-s Indium (49) --------------------- In 113m S 8X10" 4X10- 3X10-7 1 X1-3 I 7X10 ' 4X10-2 2X10-7 1X10-3 In 114m S 1X10-7 5X10-4 4X10-9 2X105 -I 2X10-8 5X10'- 7Xl10,0 2X10-s In 115m S 2Xo10- 1X10-2 8X10-8 4X10-4 I 2X10-6 1X10-2 6X!0-8 4X10-' In 115 S 2XI0-7 3X10-s 9X10- 9X10-5 I 3X10-8 3X10-3 1 X 10-9 X10 -Iodine (53) --------------------- I 126 S 8X10-' 5X 10-5 3X10-1 2X10 -I 3X10-7 3X10-s 1> 10-8 9X10-5 1 129 S 2XI0-9 1XI0-5 6X10-1t 4X10-7 I 7 X 10-a 6X 10-3 2X 10-9 2X 10-4 1 131 S 9X10-s 6XI0-5 3X10-0o 2X10 -I 3X10-7 2X 10-3 1 X 10-8 6X10-5 1 132 S 2X10-7 2X10-3 8X10-o 6X10-5 I 9X 10-7 5X10-3 3X10-6 2X10-4 1 133 S 3X10-8 2X10-4 1X10:9 7X10 -I 2X107 1 X 10-3 7 < 10-t 4X10-5 1 134 S 5X10-7 4Xl0-3 2XI10-8 1X10-4 I 3X 10- 2X 10-2 1 X 10-? 6X10 -1135 S 1X10-7 7 X 10-4 4X 10- 2X10 -I 4X 10-7 2X 10-3 1 X 10-8 7X 10-5 Iridium (77) --------------------- Ir 190 S 1X 10- 6X10-3 4X10-s 2X10 -I 4X10-7 5X10-s lXI0-8 2X10-4 Ir 192 S 1X10-7 1X10-3 4X10-' 4X10-5 I 3X10-8 1X10-3 9X10-1o 4X 10 -Ir 194 S 2X10-7 1X10-3 8X10-9 3X 10 -I 2X 10-7 )X10-4 5X 10- 3X10-5 Iron (26) ------------------------ Fe 55 S 9X10-7 2X10-2 3X10-s 8X10-4 I 1 X10 7X 10-2 3X 10- 2X 10-3 Fe 59 S. 1X10-7 2X10-3 5X10-9 6X10-5 I 5X 10-' 2X 10-3 2X 10-2 5X 10-5 Krypton 2(36). --- —-— _____ Kr 85m Sub 6X10-56. ---.....-I.X 1X1O-7 Kr 85 Sub 1X10-s - -3X10-7.............. Kr 87 Sub 1 X 10 --- 2X10-s Lanthanum (57) --- —------------- La 140 S 2X10-7 7 X I0 — 5X10-9 2X10.5 I X 10-7 7 X 10- 4X10-9 2X 10 -Lead (82),..-. —,.. Pb 203 S 3X10-6 1X10-2 9X10-8 4X10-4 I 2X10-6 1X 10-2 6X10-8 4X10- Pb 210 S 1X10-10 4X10-6 4X10-12 1X10-7 I 2X10-10 5X10-3 8X10-12 2X10-4 Pb 212 S 2X 10-8 6X10-4 6X10-l1 2X10-5 I 2X10-8 5X10-i 7XO1-lo 2X10-5 Lutetium (71) --------- ------- Lu 177 S 6X10-7 3X10-3 2X10-8 1-)<10 -I 5X10-7 3X10-3 2X10-e 1X10-4 Manganese (25) --- —- Mn 52 S 2X10-7 1X10-3 7X10-9 3X10-5 I 1X10-7 9X10-4 5X 10-9 3X10-5 Mn 54 S 4X10-7 4X10-3 1X10-o 1X10t I 4X10-' 3X10-3 1XOl-9 1X10-4 Mn 56 S 8X10-7 4X10-3 3X10-8 1X10-i I 5X10-7 3X10-3 2X10-s lX1 0 -Mercury (80). --- —--------------- Hg 197m S 7X10-7 6X10-3 3X10-8 2X1fi-4 I 8X10-7 5X 10-3 3X 0- 2Xo10 -Hg 197 S X 10-6 9X10-3 4X10-8 3X 10-t I 3X10- 1 X10-2 9X10-8 5X10-4 Hg 203 S 7X10-8 5X10-4 2X10-' 2X10-s I 1X10-7 3X10-3 4X10-9 1 X10 -Molybdenum (42) - Mo 99 S 7X10-7 5X10-a 3X 10-s 2X10-4 I 2X10-7 1X10-' 7X10-o 4X10 -Neodymium (60).....-... Nd 144 S 8X10-1 2X10-3 3X10-1s 7X10-5 I 3X 10-1o 2X1 10-0 -11 8X10-a See footnotes at end of table.

Table 1.10 Contd. Table I Table II Element (atolnic number) Isotope I Column I Column 2 Column I Column 2 Air Water Air Water (,c/nml) (pc/ml) 0(c/ml) (pc/ml) Neodymium (60) - -----— _. Nd 147 S 4XI0-7 2X10-3 1X10-S 6X10-5 I 2X10'7 2X10-3 8X10-9 6X10-5 Nd 149 S 2X10-~ 8X 10-3 6X10-s 3X10-4 -I 1X10-5 8X10-3 5X10-s 3X10-4 Neptunium (93). --- —-Np --- —-- N. N 237 S 4X10-12 9X10-5 1X10-13 X10-4 I 1 X 10- 9X 10-4 4X10-12 3X 10 -Np) 239 S 8XI0-7 4X10-3 3X10-8 1X1O-i I 7XO1-7 4X1(-3 2X10-8 1X10-4 Nickel (28) ---------------------- Ni 59 S 5XI0-7 6X10-3 2X10 — 2X10-4 I 8XI10-: 6X 10-2 3X 1-(d 2X 10 -Ni 63 S 6X 0- 8 X 10-4 2X10-9 3X10-5 I - 3X10-7 2X10-2 1X10-s 7X 10 -Ni f65 S 9X 10- 4Xl0-3 3X10-e 1> 10-' I 5X10-7 3X 10-3 2X10-, 1 X 10-4 Niobium (Columnbium) (41)- ShNb 93r S IXI10-7 IX1(-2 4X10-9 4X10-4 I 2X10-7 IX 10-2 5X10-9 4X10-4 Nb 95 S 5X1I0- 3XI0-3 2X10-8 1X 10-4 I X 10-7 3X 10-3 3X 10-' 1X10-4 Nb 97 S 6X10-6 3X10-2 2X10-; 9X10-4 I 5 X 10-6 3X 1(-2 2X 10- 9 X 10-4 Osmium ( 6) -................- Os 185 S 5X10- 2X1()-3 2XlO-e 7X 10-s I 5X 10- 2X 10-3 2XIO- 7X10-6 Os 1911rm S 2X 10- 7X 10-2 6Xi10-7 3X 10 -I 9X 10- * 7X 10-2 3Xl1-1 2X10-3 Os 191 S I 1X ()-6 5X 10-3 4X 10-8 2X10-4 I 4X 10-; 5 >10-3 I X 10- 2X10-4 Os 193 S 4 X 10-; 2X 10-3 1X 10- 6X 10-5 1 3XI0 — 2XI0-3 9X10-9 5X10-5 Palladium (46) ------------------ Pd 103 S 1X10- 1X 10-2 5X10-3 3X10-4 I 7X 1(-7 8X10-3 3XIO-s 3X10-4 Pd 109 S 6X10- 3 X10-3 2X10-s 9X10-5 I 4X 10-7 2X 10-3 X >10-8 7X10-s Phosphorus (15) ------------- P 32 S 7X10-8 5X0 -t 2X 10-9 2X 10 -I 8>X10-s 7X10-4 3X10-9 2X10-3 Platinum (78) ------------------ Pt 191 S 8X10-7 4X 10-3 3X 10-8 1X 10 -I 6X10-7 3X10-3 2X 10-8 1X10-4 Pt 193m S 7 X l)- 3X10-2 2X10-7 1X10-3 I 5X 10-6 3X 10-s 2X10-7 1 X10-3 Pt 197m S 6X10- 3X 10-2 2X 10-7 IX1-3 I 5LX 10- 3X 10-2 2X 10-7 9X10-4 Pt 197 S 8X10-7 4X10-3 3X 0-8 110-4 I 6X 1(-7 3X 10-3 2X10-8 1 X 10 -Plutonium (94) ------------------ Pu 238 S 2X1(-12 1X10-4 7X10-1" 5X10-s I 3X10-11 8X10- 1X10-12 3X 10-' Pu 239 S 2X10-12 1X10-' 6X10-l 5X10-' I 4X10-11 8XI0-4 1X10-2 3X 10-5 Pu 240 S 2X 10-12 1 X 10-4 6X 1()- 5X 10 -I 4X 10-11 8X 10-4 1X10-12 3X10-5 Pu 241 S 9X10-11 7X10-3 3X10-12 2X 10-' I 4 X 10-8 4X10 ' 1>10- 1 10-3 Pu 242 S - 2X10-12 1X 10- 6X10-l4 5X10-6 I -4X10-it 9X10-4 1 X 10-12 3X10-s Polonitim (84).. —............... Po 210 S 5X10 —o 2X10-' 2X10-ll 7X10-7 I 2X 1 0"' 8X10- 7 X 10-12 3X 10 -Potassium (19)..-................ 1K42 S. 2X10-s 9X 10-3 7X1O-8 3X10-4 I IX10-7 6X10-4 4X 10-9 2X10-5 Praseodymium (59)............... Pr 142 S 2X10-7 9X10-4 7X10-9 3X10 -I 2X10-7 9X10-4 5X10-9 3X10-' Pr 143 S 3X 10- 1X 10-3 1X 10- 5X10-6 I 2X10-7 1X10-3 6X10- 5X 10-s i'romethium (61) -..........., Pm 147 S 6X10-8 6X 0-3 2X10-9 2X10-4 I 1X 10-7 6X 10-3 3>1(0- 2 X 10-4 Pm 149 S 3X10-7 10XI-3 1 10-8 4X10-5 I 2X10-7 1 X 10-3 8X10- 4 X 10 -Protoactinium (91) -------------- Pa 230 S 2X10-9 7X 10-3 6X10-11 2X10 -I 8X 10-l 710- 1 0-113 3X 2X 10 -Pa 231 S lXI0-12 3X10-5 4X10-1' 9X10-7 I 1XI0-1o FX 10-4 4X (0-r 2X10 -

1-33 Table 1.10 Contd. Table I Table II Element (atomic number) Isotope 1 Column 1 Column 2 Column I Column 2 Air Water Air Water (Jc/ml) (pc/ml) (c/mi) (/c/ml) Protoactinium (91) ~........ Pa 233 S 6X10-7 4X10-3 2X10- 1X-10-4 I: I 2X10-7 3X 10-3 6X 10-9 1X 10 -Radium (88) --------------------- Ra 223 S 2X10-9 2X10-5 6X10-11 7X10-7 I 2X -10- ~ 1XI0-4 8X10-2 4X10" Ra 224 S 5X10-9 7X10-s 2X10' - 2X10 — I f 7X10-'0 2X10-4 2X10-1l 5X10-' Ra 226 S 3X10-" 4X10-7 1X10-12 1X10-8 I I 35X10-" 9X10-4 2X10-12 3X10-5 Ra 228 S 7X10-" 8X10-7 2X10-12 3X10-8 I 4X10-11 7X10-4 IX10 —2 3X10-5 Radon (86) ----------- ------- Rn 220 S 3X10-7 -----— X --- — 10-8 - s. Rn 222 S 1X10-7 __-_____ —____ 3X10-9 -------------- Rhenium (75) ------- ------- Re 183 S 3X10-6 2X10-2 9X10-8 6X10-4 I 2X10-7 8X10-3 5X10-9 3X10-4 Re 186 S 6X10-7 3Xl0-3 2X10-8 9X 10-5 I 2X1-7 1 X 10-3 8X 10-9 5X10-5 Re 187 S 9X10-6 7X10-2 3X10-7 3X10-3 I 5 X 10- 4 X 10-2 2 X 10-b 2 X 10-3 Re 188 S 4X10-7 2X10-3 1X10-3 6XI0-5 I 2X 10-7 9X 10-4 6X 10-9 3X 10-5 Rhodium (45) -- Rh 103m S 8X10-5 4X10-1 3X 10-6 I X 10-2 I 6X 10-56 3X 10- 2X 10-6 IX 10-2 Rh 105 S 8 X 10-7 4 X 10-3 3X 10-8 1 X 10-4 I 5X 10-7 3X 10-3 2X10-8 IX 10-4 Rubidium (37) --------- Rb 86 S 3X10-7 2X10-3 1 X 10-9 7X1(f5 I 7 X 10-s 7X10 —4 2X 10-9 2X10-5 Rb) 87 S 5X 10-7' 3X 10-3 2X 10-8 1X 10-4 I 7X 1 0-4 5X 10-3 2X 10- '2X 10-4 Ruthenium (44) --- —------------- Ru 97 S 2X 10-6 1X10-2 8X10-8 4X10-4 I 2X10-6 1X 10-2 6X10-8 3X 10-4 Ru 103 S 5X 10-7 2X 10-3 2X 10-8 8X 10-5 I 8X 10- 2X 10-3 3X10-9 8X 10-s Ru 105 S 7X1 10-7 3X10-3 2X10-9 IX 1(-4 I 5X10-7 3X10-3 2X10-s lXl 1-4 Ru 106 S 8X10-s 4X10-4 3X10-9 1Xls - I 6X 10-9 3X 10-4 2X 10-0 1X 10-5 Samarium (62) -.-. --- —------- Sm 147 S 7X10-'1:'2X10-3 2X 10-12 6X 10-5 I 3X10-10 2 X 10-3 9X10-12 7 X 10-s Sm 151 S 6X 10-8 1XIO-2 2X10-9- 4X10-4 I 1X 10-; IX 10-2 5X109- 4X10-4 Sm 153 S 5Xl-; '2X10-3 2X10-8 8X10-5 I 4X 10-7 2 X 10-3 1 X 10-8 8X 10-5 Scandium (21) -_ ___________ ---- Sc 46 S 2X 1()-7 X 10-3 8X 10-9 4X 10-5 I. 2X " 10-' 10-(3 8X10-10 4X10-5 Sc 47 S 6X 10-7 3X10-3 2X10-s 9X10-5 I 5X 10-7 3X 10-3 2X 10-e 9X 10 -Sc 48 S 2X1(0-7? XX1(4 6X10-9 3X10-5 I 1X l0-7 xx 1-4 5X 10-9 3X 10 -Selenium (34) -------------------- Se 75 S lX 10- 6 9 X 10-3 4X-10s 3X10-4 I 1X10-; 8X 10-3 4X10-o 3X10-4 Silicon (14) -------------- I ------ Si 31 S 6Xl 106 3X10-2 2X10-7 9X1O-4 I X 1()-6 6X 1-3 3X 10- 2 X 10-4 Silver (47) ---------------.- Ag 105 S 6X10 —7 3X10-3 2X10-8 1X10-4 I 8X 10-' 3X 10-3 3X10-9 1X10-4 Ag 110m S 2X10-7 9X10-4 7X10'- 3X10-5 I 1X 10-s 9X 10-4 3X10-1o 3X 10 -Ag 111 S 3XI0-7 I1X10-3 X 10-8 4X10-5 I 2X 10-7 1X 10-3 8X 10-9 4X10-5 Sodium (11) ------------------ Na 22 S 2X10-7 1X10-3 6X10-9 4X10-s I 9X 10-9 9X 10-4 3X 10-lo 3X10-5 Na 24 S 1X10-6 6X10-3 4X10-e 2 X 10 -I 1X 10-7 8X 10-4 5X 10-9 3X 10 -Strontium (38) ------------------ Sr 85m S 4Xl0-5 2X10-1 1X10-6 7X10-3 I 3X10-5 2X 10-1 1X 10- 7X 10-I Sr 85 S 2X10-7 3X10-3 8X10-9 1X10-4 I 1 X10-7 5X 10-3 4X 10-o 2X10-4 Sr 89 S 3X10 — 3X10-4 1X10-o lX10-5 I 4X 10- 8X10-4 1X10-9 3X10-5 See footnotes at end of table.

1-34 Table 1.10 Contd. Table I Table II Element (atomic number) Isotope ' Column 1 Column 2 Column I Column 2 Air Water Air Water (pc/mi) (Mc/ml) (pc/ml) (c/mml) Stronttum (38)_-________.- Sr 90 S 3X10-1o 4X10-6 X1X -11 1X10-7 I 5X10-'5 1X10-3 2X10-1o 4X10-6 Sr 91 S 4Xl0- 2X10-3 2X10- 2X10-8 7X10-5 I 3X 10-7 1X 10-3 9X10-o 5X 10 -Sr 92 S 4X10-7 2X10-3 2X10-8 7X10'I 3X 10-7 2 X 10-3 1 10-8 6X 10 -Sulfur (16) --------- ------------- S 35 S 3X10-7 2X10-3 9X10-o 6X104 -I 3X10-7 8XIO- 9X10' 3X10-' Tantalum (73) ------------------ Ta 182 S 4X10-s 1X10-3 1X10-o 4X10-' I 2X10-8 1X10-3 7X10O-1o 4X10-6 Technetium (43) ---------------- Tc 96m S 8X10-5 4X10-1 3X10-6 IX10-s I 3X10-5 3X10-' 1X10-1 1X10-2 Tc 96 S 6X10-7 3X10-3 2X 10- 1 X 10-4 I 2X10-7 1 X10-3 8X10-9 5X 10 -Tc 97m S 2X10- 1X10-2 8X10-s 4X10-4 I 2X10-7 5X10-3 5X10-' 2X10-4 Tc 97 S 1X10-5 5X10-2 4X10-7 2X10-8 I 3X10-7 2X10-2 X10- 8X10-4 Tc 99m S 4X10-s 2X10-i 1X 10-6 6X10-3 I 1X10-s X10-2 5X 10-7 3X10-3 Tc 99 S 2X10- 1X10-2 7X10(- 3X10-4 1 6X 10- 5X10-3 2X 10- 2X 10 -Telluriuni (52) -------------------- Te 125m S 4X10-7 5X1(-. lXl0-s 2XI0-4 I 1X10-7 3X10-3 4X10-' IX10-' Te 127m S 1X10-7 2X10-3 5X 10-9 6XIO-5 I 4X10-8 2X10-3 IX 1t0- 5X10-6 Te 127 S 2X10-6 8X10-3 6X10-8 3X10-' I 9X10-7 5X10-3 3X10-8 2X10-4 Te i29m S 8X10-8 IXI0-3 3X10-9 3X10-' I 3X10-8 6X10-4 1X 10-' 2X10-' Te 129 S 5X10- 2X 10-2 2X10-7 8X10-4 I 4X10- 2X10-2 1X10-7 8X10-4 Te 131m S 4X10-7 2X10-3 1X1-8 6X10-s I 2X10-7 1X10-s 6X10-' 4X10-6 Te 132 S 2X10-7 9X10-4 7X10-9 3X10-s I 1X10-7 6X10-, 4X10-o 2X10-' Terbium (65) ------------------- Tb 160 S 1X10-7 1X X0-3 3X10-o 4X10-s I 3X10-s 1 X 10- lX lo0- 4X10-s Thallium (81) ------------------- T1 200 S 3X10- 1X10-2 9X10-s 4X10-4 I lX10-6 7X10-3 4X10-8 2X10-4 T1 201 S 2X10-6 9X10-3 7X10-8 3X10-'4 I 9X10-7 5X 10-3 3X10-8 2X10-' TI 202 S 8X10-7 4X10-3 3X10-8 1X1O-4 I 2X10-7 2X 10-3 8X 10- 7X10'T1 204 S 6X10-7 3X10-3 2X 10-8 IX10-4 I 3X10-8 2X10-3 9X 10-0 6X10'Thorium (90) -------------------- Th 228 S 9X10-2 2X10-4 3X10-13 7X10-4 I 6X10-12 4X 10- 2X10-13 10' Th 230 S 2X 10-2 5X10-5 8X10-14 2X10-4 I 10-l1 9X 10-4 3X 10-3 3X10-' Th 232 S 3X10-11 5X10- 10-12 2X104 I 3X10-"11 M3 10-12 4X10-' Tb natural S 3X10-11 3X10-' 10-1 10-4 I 3X10-" 3X10"-4 10-1 10-& Th 234 S 6X 10-8 5X10-4 2X10-9 2X10-m I 3X 10- 5X10-4 10-4 2X 10-' Thulium (69) --------------------- Tm 170 5 4X10-s 1X 10-' 1X10-o 5X10-I I 3X 10- 1X10-3 1X10-o 5X10-' Tm 171 S IX10-7 1X10 -2.X10-o 5X10-4 I 2X10-7 1 X 10-2 8X 10- 5X10-4 Tin (50) ------------------------ Sn 113 S 4X10-7 2X1X13 1X0 —8 9X10-' I 5X10-' 2X10-3 2X10-' 8X10-A Sn 125 S 1X0-7 5X 10-4 4X 10-o 2X10-6 I 8X10-8 5X10-4 3X 10- 2X10 -Tungsten (Wolfram) (74) -------- W 181 S 2X10-6 1Xl0- 8X10-8 4X10-' I IXl0-7 1X10-2 4XlC-' 3X10-4 W 18l S 8X10-7 4X1O-3 3X10-s 1X10-4 I 1 X10-7 3X10-3 4X10'-o lX10-4

1-35 Table 1.10 Contd. Table I Table II Element (atomic number) Isotope l Column 1 Column 2 Column i Column 2 Air Water Air Water (Pc/ml) (pc/ml) (Pc/ml) (Jc/ml) Tungsten (Wolfram) (74). --- —-- W 187 S 4X10-7 2X10-1 2X10-' 7X10-3 I 3X10-7 2X10s 1Xl1O- 6Xi10 -Uranium (92) ----------------- U 230 S 3X10-10 1X10- IXlO-"l 5X10oI IXi10-1 lX 10-4 4X10-1 5X10 -IJ 232 S 1XlO-1O 8X1O-4 3XiO1-I 3X10-s I 3X10-1l 8X10-4 9X10- 13 3X10-s U 233 S 5X1"0lo 9X10I4 2X10-11 3X10-5 I 1X10-1o 9X10-4 4X10-U 3X1-s5 U 234 8 6X10-lo 9X10-' 2X10-11 3X106 -I 1X 10-o 9XI10- 4X10'-1 3X105 U 235 S 5X100-1 8X10-4 2X10-l 3X10-' I IX10-lo 8X10-4 4X10-12 3X104 U 236 S 6X10-i0 1X10-3 2X10-11 3X10-5 I lXO-10-o lX10- 4X10"2 3X10-s U 238: S 7X10ll l X10-3 3X10-2 4X10-5 I 1 X10-i 1X0-t 5X10-!2 4X105 -U-natural S 7X10-1 5X10-' 3XlO-t 2X10-5 I 6X10-11 5X10-4 2X10-12 2X10-5 Vanadium (23). --- —------------- V 48 S 2X 10- 9X10-4 6X10-e 3X10-5 r 6X10-8 8X 10- 2X10-9 3X10rXenon (54) --------------- Xe 131m Sub 2X1O- - -4X10-7 Xe 133 Sub 1XlO-5 -_-_._ ---_-_.- 3X10-7 ------------- Xe 135 Sub 4X10-6 1X —0-7 Ytterbium (70) ------------------ Yb 175 S 7X10-7 3X1-'t 2X10-8 1X10-4 I 6X10-7 3X10-3 2X10-6 1X 1-4 Yttrium (39) Y 90 -, 1Xl1g-7t 6X10-4 4X10-9 2X10-5 I 1X10-7 6X10-4 3X1O-9 2X10 -Y 91m S 2X10-5 1XIO-I 8X10-t 3X10-3 I 2X10- 1 X 10-1 6X10-7 3X 10 -Y 91 S 4X10-s. 8X10-4 1X10- 3X10-5 I 3X10-'- 8X 10-4 1-X10-9 3X10-5 Y 92 S 4X10-7 2X10-3 1XI0-6 6X10-5 I 3X0-7 2 2X 10-3: X 10-8 6X10-5 Y 93 S 2X10-7 8X10-' 6X10-' 3X10OI 1X1O,? 8X10-4 5X10-' 3X10-5 Zinc (30) --- —------------------ Zn 65 S 1X10-7 3X10-3 4Xl-9 iX10-l I 6X 10- 5X 10- 2X 10- 2X 10 -Zn 69m S 4X10-7 2X10-3 1X10-8 7X10-s I 3X 10-7 2X 1-? 1X 10 — 6X10-5 Zn 69 S 7X10 5X10- 2X10-7 2X10-3 I 9X10-4 5X10-2 3X10-7 2X 10 -Zirconium (40) --------------- Zr 93 S 1X 10-7 2X 10- 4XIr0- 8X10-4 I 3X110-7 2X10-2 1X10-8 8X10-4 Zr 95 8 1X10-7, 2X10- 4X 10-9 6X10-5 I 3X 10- 2X10- 1X 10-9 6X 10-5 Zr 97 S I X 10-7 5X10-4 4X10-0 2X10-6 I 9X10-s 5X10-4 3X10-o 2X10-5 S Soluble (8); Insoluble (I). MPC's, are MPCA, and MPCB, and MPCc respec2"Sub" means that values given are for submersion tively, then the concentrations shall be limited so that in an infnite cloud of gaseous material. the'following relationship exists: NOTr: In any case where there is a mixture in air or water of more than one radionuclide, the limiting values CA + C B Cc for purposes of this Appendix should be determined as MPCA MPCB MPCC follows: 1. If the identity and concentration of each radionu- 2. If either the identity or the concentration of any clide in the mixture are known, the limiting values radionuclide in the mixture is not known, the limiting should be derived as follows: Determine, for each ra- values for purposes of Appendix B shall be: dionuclide in the mixture, the ratio between the quantity a. For purposes of Table I, Col. 1-1 X 10-12 present in the mixture and the limit otherwise estab- b. For purposes of Table I, Col. 2-3X10-7 lished in Appendix B for the specific radionuclide when c. For purposes of Table II, Col. 1-4Xl0-14 not in a mixture. The sum of such ratios for all the d. For purposes of Table II, Col. 2-1X1Oradionuclides in the mixture may not exceed "1" (i.e., 3. If the conditions specified below are met, the cor"unity"). - responding values specified below may be used in lieu EXAMPLE: If radionuclides A, B, and C are present of those specified in paragralh 2 above. in concentrations Ci, CB, and Cc, and if the applicable

1-36 As of January 1, 1961 revised AEC regulations went into effect, (i1yf7):, These new regulations differ only slightly from the NCRP and ICRP recommendations, but for purposes of comparison will be discussed in the next section. It should be noted that while the ICRP and NCRP may make recommendations, these organizations have no legal authority in so far as regulation and control of radioactive substances is concerned. In summary, the men responsible for the recommendations made by ICRP are experts in physics, biology, genetics and radiation protection. They have made conservative and intelligent estimates of permissible dosages, based upon sparse data on human radiation injury, and extrapolation from extensive animal experimentation. As more detailed information about radiation effects is accumulated, it may be anticipated that these permissible levels will be further altered. 1.23 Regulation and control by AEC Before 1946 all regulation and control of radioactive material where it existed was based on local and state regulations, and to a lesser extent, the federal government in so far as interstate commerce was 'concerned. In 1946, however, the Atomic Energy Act (Public Law 703, 83rd. Cong., 2nd. Sess., 60 Stat. 919) was passed. This was followed by the Atomic Energy Act of 1954, which was an amendment to the original act. The act as summarized by the McMahon Committee was to initiate the following~o (1) An Atomic Energy Commission whose members should be appointed by the President, with the advice and consent of the Senate. (2) Control by the Commission over all atomic energy installations, with the authority to acquire source materials,, i.e. uranium ores, and other property needed in the development of atomic energy (3) Encouragement by the Commission of research, development, and exploitation in the field of atomic energy with the power to license property and facilities for these purposes. (4) A prohibition on the export or import of source materials except under the direction of the Commissionr. (5) Security regulations and penalties for their violation to be prescribed by the Commissiono The AEC thus was given in the act (among other things) the duty to issue licenses to individuals and organizations for the use of byproduct material, Byproduct material is any radioactive material (except special nuclear material) produced by a nuclear reactor. The AECG, therefore, has

1-37 control only in so far as issuance of licenses are concerned. In actual practice, however, this control covers practically all of non-naturally occurring radioactive material since there are few nuclear accelerators large enough to produce large amounts of active material. In order to maintain a license, once it has been issued by the AEC, the licensee must comply with the regulations as written in the Federal Register (Title 10, Part 20). (107) The regulations are in general a restatement of the recommendaations of the ICRP and the NCRP with a few exceptions. The most important differences are given in the following section. 1.24 Differences between AEC regulations and ICRP recommendations (1) The AEC regulations do not apply to x-radiation as produced by any mechanical device not containing byproduct radioactive material. X-ray machines are in general licensed by the state or local governments, (2) The AEC regulations do not apply to any naturally occurring radioactive nuclides such as radium and its daughters if the substance was not produced in a nuclear reactor. (3) The AEC regulations state the appropriate radiation signs and symbols to be used in radiation areas and in regard to labeling source and byproduct material~ (4) The AEC regualtions describe what records must be kept and detail when a licensee must make reports to the AEC regarding radiation incidents (5) The AEC regulations have set the following exposure limits for individuals in restricted areas: A. No individual shall possess, use, or transfer, radioactive material in such a manner as to cause any person in any period of one calendar quarter to receive a dose in excess of the limits specified in Table 1.11

1-38 Table 1.11 Maximum Permissible Doses in Rems per Calendar Quarter (107) 1. Whole body; head -and trunk; active blood-forming organsI lens of eyes; or gonads -- -------------— 1 1/4 2. Hands and forearms; feet and ankles --- —-------------------------— 18 3/4 3. Skin of whole body -------------------------------------— _ ---7 1/2 B. An individual may receive a dose to the whole body greater than that permitted in paragraph (1) A provided: (i) That during any calendar quarter the dose to the whole body does not exceed 3 rems; and (ii) That the dose to the whole body when added to the accumulated lifetime dose does not exceed 5 (N-18) rems, where "N" equals the individual's age in years; and (iii) That "dose to the whole body" shall include any dose to the whole body, gonads, active blood-forming organs, head and trunk, or lens of the eye. (6) Exposure of minors —The regulations prohibit a licensee from causing an individual within a restricted area, who is under 18 years of age, to receive a dose in excess of 10 per cent of the limits specified in (5) above. 1.25 Practical limits-comparison with "background" Sufficiently sensitive radiation detection apparatus invariably records the presence of radiation, even in the total absence Qoff known adiatin.s*o rces, This omnipresent radiation level, which may exhibit rather wide fluctuations with time and differs significantly from place to place, is called "background radiation." Because every living organism is subjected to this radiation throughout life, an evaluation of the dosage received and its likely effects upon existing species is essential to an intelligent selection of maximum permissible radiation exposure levels. The major portion of total background radiation levels (roeughly 70 per cent at sea level) can be shown to be due to the bombardment of the earth by high-energy particles from outer space. These particles are considered to be

1-39 predominately protons and are called "cosmic rays." This radiation is known to remain fairly constant at any one location, but to demonstrate systematic variations with altitude and latidude. The remaining fraction of total background radiation levels, which demonstrates wide local fluctuations with time and location, is due to naturally occurring radioactive substances. In addition to the four radioactive series of heavy elements (the uranium, thorium, actinium, and neptunium series) seven other radioactive isotopes occur in nature. These are listed in Table 1.12. (88) Table 1.12 Natural Radioisotopes Other than Heavy Elements (88) Isotope % of Half life Radiation natural element alpha beta gamma 40 9 K.0119 1.3 x 10y x x Rb87 27.2 6.3 x 10 y x x 124~~~~ 12 Sn24 6.0 6 x 1013y x Nd154 1 x y x 101 Sm59.26.6 2 x 10y x Lu176 2.5- 2.4 x 10oy x x Re187 62.9 4 x 10 y x The naturally radioactive materials are present in the soil, air and water and are present in measurable quantity in uhe human body. For example, a layer of typical soil, one foot thick and one square mile in area, contains 6 tons of uranium and 12 tons of thorium. Among the decay products of the uranium and thorium in the ground-is the radioactive gases radon and thoron. Radon and thoron decay, in turn, to a radioactive charged particles which for the most phart attache to dust particles. It is the radon and thoron always present in the air which contributes to the wide, nonstatistical variations of background radiation.

1-40 Total background radiation varies from 0.01 to 041 mr/hr, varying with location. This constitutes a yearly dose of from 0.09 o 0.9.r for everyone. Concentrations of uranium and radium in drinking water may vary from the maximum permissible concentrations listed in Table 1o1.O should serve to emphasize how conservative ICRP has been. Considerable data on natural background radiation levels have been accumulated and surveys on radiation levels are being continued at many places throughout the world. It will thus be possible, as the use of nuclear energy increases, to evaluate the effect of this program on the radiation background. These survey points are also very useful in evaluating the local effect, on both radiation and contamination levels, of specific nuclear energy facilities. 1.26 Comparison with nonoccupational exposures The maximum permissible exposure levels discussed in the previous sections were established to apply only to persons aware that they are working in the field of radiation. Permissible levels of radiation and the quantities of radioactive materials released in areas accessible to the general public should be much lower. In addition to the occupational exposures accumulated by persons working with radiation and continuous exposures to background radf iation experienced by everyone, important radiation exposures are occasional ly sustained by laymen from other sources. Radiation levels as high as 10 mr/hr have been obseve b oeed at the face of a radium-dial watch. At such a dosage rate, the weekly permissible dose of 100 mr could be accumulated in a mere 10 hours. The fact that this dose is very localized when received, and is received only intermittently, substann tially reduces the net hazard. There is no record of any serious external radiation injury from radium watches. X-ray and fluoroscopic examination result in occasional radiation exposures considerably in excess of permissible levels, both to individuals employed with radiation and to the general publico Figure 1o7 illustrates some common X-ray exposures. Notice the dosage received during fluoroscopic examination. An inepetrienced or over-enthusiastic fluoroscope operator could easily deliver a damaging dose of radiation during a prolonged exposure (89)oPermanent total sterility of the human male or female requires dosages of from 400 to 1000 rep. to the gonads. As such exposures would usually be

1-41 CHEST (POSTERIOR-ANTERIOR) CHEST (POSTERIOR - ANTERIOR, PHOTOFLUOROSCOPIC X-RAY) LUMBAR, SPINE (ANTERIOR-POSTERIOR) LUMBAR, SPINE (LATERAL) PELVIS PREGNANCY (ANTERIOR-POSTERIOR) PREGNANCY (LATERAL) KIDNEY-URETER- BLADDER ABDOMEN HEART DISEASE (INCL. FLUOROSCOPY AND CATHETERIZATION) / / TO 140r Y///////ZA GASTROINTESTINAL SERIES (TOTAL OF 6 FILMS) GALL BLADDER EXTREMITIES SKULL (POSTERIOR-ANTERIOR) DENTAL FILM AVERAGE FLUOROSCOPIC EXAMINATION 10 T020 r/min //// 0t 1 2 3 4 5 6 7 8 9 10 11 O. ROENTGENS Figure 1.7 Exposures from common X-ray examinations (17) Figure 1.8 Engine crew moving car loaded with highly radioactive waste to Hanford burial site. Note spacing of several cars to provide protection through distance (Courtesy Genaral Electric Co.)

1-42 lethal if received by the whole body, the common fear of being made sterile by radiation exposure is largely academic. However, reduced fertility has been reported at lower levels of exposure (8). As more and more people become aware of the problems of radiation expo'sure control, bet er records will be kept of cumulative individual radiation exposures,- and greater care will be taken to minimize total exposure during necessary medical procedures. 1.27 Three safety rules for protection from external radiation Minimization of personal injury from external radiation can be accomplished by judicious use of three simple rules involving: distance, time, and mass. Distance A radioactive material emits radiation in all directions. As the distance between the worker and the radioactive material is increased, the exposure rate decreases. For a point source of gamma radiation this rate of decrease is inversely proportional to the square of the distance. Thus, the first and simplest rule for protection of personnel exposed to external radiation hazard is: "Stay as far away from the radiation source as possible while working with the source or in a radiation area." In laboratory work the use of tongs and long-handled toqls is preferable to hand manipulation. Remote control is preferable to the use of tongs. When large amounts of radioactive material must be handled without shielding, special techniques are used. Figures 1.8, and 1.9. shbw techniques used in the burial of highly radioactive equipment at the Hanford burial site. The equipment is first prepared for removal and burial behind heavily shielded walls called "canyons" with the use of remote manipulators. The material is removed from the canyon and taken to the burial site by railroad using a train of several cars to provide sufficient distance between the engine crew and the source of radiation (see Figure l b 8)o The car is lifted from the track by a long-boom crane shown in the upper left of Figure 1o 9. and dragged by a long cable to a pit previously excavated. Figure 1loI9 also shows the radiation field being monitored with a "Cutie-Pie" type survey meter by the health physicist employee shown at the right. Earth will be pushed into the pit by use of bulldosers with metal shields for protection of the operators.

S...B.......r.......f..s.B......r....f.......:y..B>..fB.rgr....r.>..-..gg.... iE;.Rr...:..r.rE-:... -..r:...... f IL- )4~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ 3~~.... Figure 1.9 Burial of highly radioactive waste at Hanford (Courtesy of General Electric Co.) Figure 1 10 Operator working behind heavy shield and using remote manipulator (Courtesy General Electric Co.) g > s. as.>2.g<B>Xr g 4. gB g.B g g 4....................................:0:: iiii ii24i i ii gg:l..........................................4. 2;.. ot.> 22. - ~- > 2;.t --- 3e >;t. f. @-<@r~~~~~~~~~~~~~~~~~~~~~~~~~~~~g~~g:: —..................... @...... l i 4g g22 gsB@V.gg g.@g; U 2;; t = gh @S gPB @.;g~~~~~~~~~~~~~~~~~~~~~~~gg;. @.@E gg 22 @ g 0 00;;0000 tl; ' t................. F S.; >. @g @.g 2 @g g @g @g@E g~g@ @Eg g g @ E @:: it t:0 t:0000 t:;0t iititii:: t..............@...........@.....:-: g 2 g g @g2BiggB@@gig2@-$g gg g2@g gjgij~~~~~ji g ig~g @,.jB,g~~g gg~........ -....................................;rgli;.S.-Ss: jj;,;. Za~~~esr {S& ygvE 8;.....t..8i...;iar;SS~ E ~.i;g; @gi.@;S..; j~~~~~~~gj,@. jig @B s@;gg @ 4B; S@...@>; S @j &,.S.S..f iggfllyi gf ji~~~~~~~~~~~~~~~~~............ t~~~sB@2i:;8 g 4>&:..:;'j @jjB~~~~~~~~~~~~~~~~~gH,,@.,j:Z..ts..t>.,8@.fi"H@>:Q8@ig;B@2.i@ g;;g-0i8 &iii&& iii ig0. igii S~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.......,S...-.,:'BBB@, 4 m,,4..i; @..,., 8sB~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~g g@ g}8Bii~~~~~~~~~gg~~~g@8 @ B & j~~~~~~~~~j@':.8 4:s::-8ig-, i..:.:-S:.:-:,Sjj~~~~~~~~~~~~~~~~~~~~~~~~~~~~i~~~i-ig-i............ _ i ';8.-g; @@8...-.s.Z~~~~~~~~~~~~~~~~~~~~~~~~~~~i~~~g~~~i~~~r~~~~fhjf0;j~~~~~~~~~sii.8ggmsXV — 8U~~~~~~~~~~~~~~~~~~~i. iss. l-" —)- i"~~~~~~~~~~~~~......,:.,.g................................ m.,:,:<. -E --:,li,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.......... Figure 1.9 Burial of highly radioactive waste at Hanford (Courtesy of General Electric C-.Figure 1.10 Operator working behind heavy shield and using remot.e.. manpulto (CuteyGnea.EeticC.

1-44 Time When the task to be performed with radioactive materials is such that high dosage rates are found where the workers must be during the operation, personal exposure may be minimized by reducing the time spent in the radiation field to a minimum. A second rule might be stated: "Leave- the radiation area as soon as possible." Repeated "cold runs" prior to an experiment with radioactivity provide the best assurance that the experiment can proceed safely and quickly when the radioactivity arrives and the "hot run" proceeds. In this way, mechanical and manual operations which may cause unexpected delay are discovered and.. corrected with no radiation hazard to personnel and the "hot run" can proceed with minimal delays. Of prime importance, also, in minimizing delays during operations with radioactive materials is the elimination or simplification of complex or intricate assembly stepso Close mechanical tolerances and squeeze fits of mechanical equipment and parts should be eliminated wherever possibble to prevent binding or jamming during a transfer. Mass When the use of distance and time are not sufficient to reduce exposures to a tolerable level, the use of shielding materials becomes necessary. As has been previously discussed, radiation interacts with matter and thus intensities may be reduced. The placing of mass (water, concrete, lead) between the radioactive:...material and the worker results in reduced radiation levels. The third rule might stateo "When it is necessary to remain near a source of radiation for an extended period, be certain that there is sufficient shielding material." For a point source, these qualitative rules may be stated mathematically for a particular physical configuration, R = Kfe dt r where R - dose received K = a constant determined by the nature of the source and the absorber, including build-up factors. t = exposure time, = linear absorption coefficient of shielding material x = thickness of absorber r = distance from radiation source It is clear that, in order to minimize R, r should be maximized (the first rule), t should be minimized (the second rule) and x should be maximized (the third rule).

The concept of a permissible dose was introduced previously. If, however, one accepts a certain dose as being "acceptable" or "safe" when taken for a certain (.limited) time then the further reduction of R, once this level has been reached, is unnecessary~ The use of shielding in addition ta. distance to protect personnel is illustrated in Figure 1.10 which shows an operator at Hanford working behind a heavy shield, using a remote-manipulation toal. The observation opening is of the lead-glass shielded-window type. The operator is using the mechanical hand to pour a liquid from one container to another and controls the procedure from behind the protective wall. Additional information on shielding is given in Chapters 2 and 4 and on health physics in References 108-118.'

1-46 References - Chapter 1 1. Morgan, K. Z.]} "Health Physics and Radiation Protection," in Medical Physics, Vol. II, Chicago Yearbook Publishers, Inc., Chicago, Ill., -1950 2. Braasch, NO K. and Nielson, M. Jol A Study of the Hands of Radiologists," Radiology, 51, 719, 1948 3. Robinson, J. N. and Engle, E. T.l "The Effect of Neutron Radiation on the Human Testes: A Case Report," J. Urol., 61, 1949 4. Frieben, J., "Cancroid des rechten Handmekens," Dtsch. med. Wschr., 28, 335, 1902 5. Becquerel, H. and Curie, P., "Action Physiologique des rayons radium," C. R. Acad. Sci., Paris, 132, 1289, 1901 6. Curie, E., "Madame Curie," Translated by V. Sheean. Wm. Heinemann Ltd., London and Toronto, 1938 7. "Control of Radiation Hazards in the Atomic Energy Program," U.S. Atomic Energy Commission, Wash., D.C., 1950 8. "Concepts of Radiological Health," U. iS.Dept. of Health, Educ. and Welfare, U.S. Govt. Print. Office, Wash., D.Co, 1954 9. Martland, HS., "The Occurrence of Malignancy in Radioactive Persona: A General Review of Data Gathered in the Study of Radium Dial Painters with Special Reference to Occurrence of Osteogenic Sarcoma and Interrelationship of Certain Blood Diseases," Amer. J. Cancer, 15, 2435, 1931 10. Code, S., "Malignant Disease and its Treatment by Radium, Williams and Wil11o Hempelmaun, L. H.o Sisco, H. and Hoffman, J. G., "The Acute Radiation Syndrome: A Study of Nine Cases and a Review of the Problem," Ann. intern. Med., 36, 1952 12. "Radiation Safety and Major Activities in the Atomic Energy Programs, July-December 1956," U.S. Atomic Energy Commission, W&sh., D.C., January 1957, ibid, Jan.-Dec. 1959, USAEC, Wash., D.C,, Jan, 1960 13. Morgan, K. Zo, "Historical Sketch of Radiation Protection Experience and Increasing Scope of Radiation Protection Problems," in Lectures at Inservice Training Course in Radiological Health," The University of Michigan, Ann:Arobr, Micho, 1951

1-47 References Chapter 1 (Contd.) i4. "Report of the International Commission on Radiological Units and Measurements (ICRU), 1956," National Bureau of Standards Handbook 62, April 10, 1957 15. Tischer, R. Go and Kurtz, G, W., "Mechanism of Action of IQnizing Radiations on Living Matter," U.S. Army Quartermaster Corps., U.S Dept. Comm., OIf. of Tech. Services, Wash. D.C.7 178, 1957 16. Title 10 - Atomic Energy, Chapter 1 - Atomic Energy Commission, Part 20 - Standards for Protection against Radiation. Federal Register, January 29, 1957 17. Plough, H. H., "Radiation Tolerances and Genetic Effects," Nucleonics, 10, No. 8, 16, 1952 18. Hine, G. J. and Brownell, G. L., Radiation Dosimetry, Academic Press, New Y ork, New York, 1956 19. Warren, S. and Bowers, J"l.Z, "The Acute Radiation Syndrome in Man," KAn. intern. Med., 32, 1950 20. Wolf, B, S., "Medical Aspects of Radiation Safety," Nucleonics, 3, 25, 1948 21.0 Lawrence, J. and Hamilton, J., '"Avances in Biological and Medical Physics," Academic Press, Inc., N. Y., 1951 22. Henshaw, P. and Hawkins, J., "Incidence of Leukemia in Physicians," J. Nat. Cancer Inst., 4, 339, 1944 23. March, H., "Leutemia in Radiologists in a 20-year period," Amer. J. med. Sci,,220, i950 24. Auerbach, 0., Friedman, M., Weiss, L.,, and Amory, H. I., 'Extraskeletai Osteogenic Sarcoma Arising in Irradiated Tissue," Cancer, N. Y., 4, 1095, 1951 25. De Young, R., "The Development of Sarcoma in Bone Subjected to Irradiation," Amer. Surgeon, l313 816, 1952 26. Evans, R. D., "Quantitative Aspects of Radiation Carcinogenesis in Humans," Acta Un. int. Cancr., 6, 1229, 1952

1-48 References Chapter 1 (Contd.) 27. Furth, J. and Upton, A. C., "Vertebrate Radiobiology: Histopathology and Carcinogenesis," Annu. Revo Nuc. Scio 3, 303, 1953 28. Goldberg, R. C. and Chaikoff, I. Lo, "Induction of Thyroid Cancer in the Rat by Radioactive Iodine," Arch. Path. (Lab. Med.) 53, 22, 1952 29. Spitz, S. and Higginbotham, NJ Lo, "Osteogenic-Sarcoma Following Prophylactic Roentgen-ray Therapy," Cancer, N. Y., 4, 1107, 1951 30. Evans, T. C., "Biological and Medical Effect of R diation at Low Levels," in Lectures at Inservice Training Course in Radiological Health, the University of Michigan, Ann Arbor, Mich., 1951 31. Yockey, H. P., "Radiation Aging and its Relation to the Principles of Health Physics," Health Physics, Vol. I, No. 4, 417, 1959 32. Chital, C. M., "Studies on the Dynamics of Morphogenesis and Inheritance in Experimental Reproduction," Jo expo Zool., 14, 1913 33. Jacobson, L. 00 and Marks, E. K., "The Hematological Effects of Ionizing Radiations in the Tolerance Range," Radiology, 49, 286, 1947 34. Burstone, M. S., "Radiobiology of the Oral Tissues," J. Amer. Dent. Ass. 47, No. 6, 630, 1953 35. Burstone, Mo S., "Studies on Effect of Radioactive Colloidal Gold on the Development of the Oral Structures of the Mouse," Arch- Path. (Lab. Med.), 50, 419, 1950 36. Ferguson, J. H., Andrews, Go A. and Brucer, M., "Blood-clotting Studies on Dogs Internally Irradiated with Radiogold,"Proc. Soc. exp. Biol., N.Y., 80, 541, 1952 37. Furth, J., Andrews, G. A., Storey, R. H. and Wish, L., "The Effect of X-irradiation on Erythrogenesis, Plasma and Cell Volumes," Sth. med. J. Nashville, 44, 85, 1951 38. Goldie, H., Tarleton, Go J., Jr., Jeffries, B. R. and Hahn, P. F., "Effect of Repeated Doses of External and Internal Irradiation on Structure of the Spleen," Proc. Soc. expo Biol., N.Yo, 82, 395, 1953

1-49 References Chapter 1 (Contd.) 39. Odeblad, E., "A Study of the Short-time Effects on the Mouse Ovary of Internal Irradiation with P-,32," Acta radiol., Stockh., 38, 33, 1952 40. Taymor, M. L., Gold, No, Sturgis, S. H., Meigs, J. V. and MacMillan, J., "Effects of Irradiation Upon the Uptake of Labeled Phosphorus in Human Carcinoma of the Cervix," Cancer, N.Y., 5, 469, 1952 41. Wachowski, T. J. and Chenault, H., "Degenerative Effects of Large Doses of Roentgen Rays on the Human Brain," Radiology, 45, 227, 1945 42. Warren, So, Holt, M. W. and Sommers, SO C., "Some Early Nuclear Effects of Ionizing Radiation," Proc. Soc. exp. Biol., N.Y., 77, 288, 1951 43. Watts W. E. and Mathieson, D. R., "Studies on Lymphocytes from Persons Treated with Radioactive Iodine," J. Lab. Clin. Med., 35, 885, 1950 44. Edelmann, A., "AAAS Symposium on Radiobiology," Nucleonics, 8, No. 4, 1951 45 Lampe, I.. and Hodges, F. J., "Differential Tissue Response to Neutron and Roentgen R diation," Radiology, 41, 1943 46. Storer, JO and Harris, P., "Incidence of Lens Opacities in Mice Exposed to X-rays and Thermal Neutrons," LA-1455, Los Alamos Scientific Labo, U.S. Atomic Energy Commission, Wash., D.C., 1952 47. Brennan, J. T. et al., "The Biological Effectiveness of Thermal Neutrons on Mice," LA-1408, Los Alamos Scientific Lab., UoS. Atomic Energy Commission, Wash., D.C., 1952 48. Beck, J. S, and Meissner, W. A., "Radiation Effects of the Atomic Bomb Among the Natives of Nagasaki, Kyushu," Amer. J. cdin. Path., 16, 586, 1948 49. Bennett, L. R., Chastain, S. Mo, Flint, J. S., Hansen, R. A. and Lewis, A. E., "The Late Effects of Roentgen Irradiation, I. Studies on Rats Irradiated Under Anoxic Anoxia," Radiology, 61, 441, 1953

References Chapter 1 (Contd.) 50. Bowers, J. Z. and Scott, Ko G., "Distribution and Excretion of Electrolytes After Acute Whole-body Irradiation Injury, I. Studies and Radiopotassium," Proco Soc. exp. Biol., N.Y., 78, 645, 1951 51. Brues, A. M,, Stroud, Ao No and Rietz, Lo, "Toxicity of Tritium Oxide to Mice," Proc. Soco exp. Biol., N.Y, 79, 174, 1952 52. Conger, A. D. and Giles, No Ho, Jr., "The Cytogenetic Effect of Slow Neutrons," Genetics, 35, 397, 1950 53. DeCoursey, E., "Human Pathological Anatomy of Ionizing Radiation Effects of the Atomic Bomb Explosions," Military Surgeon, 102, 427, 1948 540 Evans, R. D., "Quantitative Inferences Concerning the Genetic Effects of Radiation on Human Beings," Science, 109, '299, 1949 55. Graff, W. S., Scott, K. G. and Lawrence, J. H., "Histological Effects of Radiophosphorus on Normal and Lymphomatous Mice," Amer. Jo Roentenolo, 55, 1946 56. Hale, W. M. and Stoner, Ro D., "The Effect of Cobalt 60 Gamma Radiation on Antibody Formation and Immunity," International Record of Medicine, 165, 358, 1952 57. Hemplemann, L. H. et al,, "The Acute Radiation Syndrome: A Study of Nine Cases and a Review of the Problem," Anno intern. Medo, 36, 279, 1952 -589 Hennessey, T. G. and Huff, R. Lo, "Depression of Tracer Ion Uptake HenesyT G adHu f. Curve in Rat Erythrocytes Following Total Body X-Irradiation," Proc. Soc. exp. Biol., N.Y., 73, 436, 1950 59. Holt, M. W., Sommers, SoCo and Warren, So, "Intra-nuclear Changes Resulting from Exposure to Ionizing Radiation as Detected in Frozendried Preparations," Lab. Investigation, 2, 408, 1953 60. Huff, R. Lo et al., "Tracer Iron Distribution Studies in Irradiated Rats with Lead-shielded Spleens," Jo Lab. clin. Med., 3, 40, 1950

1-51 References Chapter 1 (Contd.) 61. Hursh, J. B., Van Valkenburg, P. A. and Mohney, J. B., "Effect of Roentgen Radiation on Thyroid Function in Rats." Radiology, 57, 411, 1951 62. Koletsky, S. and Christie, J. Ho, "Effect of Antibiotic on Mortality from:Internal Radiation," Proc. Soc. exp Biol., N.Y., 75, 363, 1950 63. Lavik, P. S., Leonards, J, R., Buckaloo, G. W., Heisler, C. and Friedell, H. L., "Ineffectiveness of In Vivo Dialysis in Prolonging Life in X-irradiated Dogs," Proc. Soc. exp. Biol., N.Y., 83, 618, 1953 64. MacIntyre, W. Jo, Friedell, H. L. and Berg, M., "The Influence of X-Irradiation on the Disappearance of Radioactive Tracers: from Circulating Blood," NYO-4014, Western Reserve University Report No. 110, U.X. Atomic Energy Commission, Wash., D.C., 1952 65. Martland, H. S., "The Occurrence of Malignancy in Radioactive Persons: A General Review of Data Gathered in the Study of Radium Dial Painters with Special Reference to Occurrence of Osteogenic Sarcoma and Interrelationship of Certain BlOod Diseases," Amer. J. Cancer, 15, 2435, 1931 66. Mole, R. i., "Whole Body Irradiation -- Radiobiology or Medicine?" Brit. J. Radiol., N. S., 26, 234, 1953 67. Noonan, T. R. and Noonan, A. M., "Effect of Roentgen Irradiation Upon Growth and Peripheral Blood Cell Levels of the Albino Rat," Fed. Proc., 11, 114, 1952 68. Patt, H. M,, "Protective Mechanisms in Ionizing Radiation Injury,'" Physiol. Rev, 33, 35, 1953 69. Rugh, R., "Radiobiology Irradiation Lethality and Protection," Military Surgeon, 112, 395, 1953 70. Soberman, R. J., Keating, R. P. and Maxwell, R. D., "Effect of Acute Whole-body X-irradiation Upon Water and Electrolyte Balance," Amer. J. Physiol., 164, 450,; 1951

1-52 References Chapter 1 (Contd.) 71. Supplee, H., Hauschildt, J. D. and Entenman, C., "Plasma Proteins and Plasma Volume in Rats Following Total-body-X-irradiation," Amer. J. Physiol., 169, 483, 1952 72. Thomson, JO F. Tourtellotte, W. W., Carttar, M. S., Cox, R. S., Jr. and Wilson, J. Eo, "Studies on the Effects of Continuous Exposure of Animals to Gamma Radiation from Cobalt 60 Plane Sources," Amer. J. Roentgenol., 69, 830, 1953 73. Trum, B. F., Haley, To J., Bassin, M., Heglin, Jo and Rust, J. H., "Effect of 400 Fractional Whole Body-irradiation in the Burro, (Equus asinus asinus)," Amer. Jo Physiol., 174, 57, 1953 74. Walso, C. Eo, "Toxicity of Inhaled or Ingested Radioactive Products," Nucleonics, 3, 1948 75. Copp, Do H. et al., "The Deposition of Radioactive Metals in Bone as a Potential Health Hazard,l! Amero J. Roentgenol., 58, 1947 76. Code, S., "Malignant Disease and its Treatment by Radium," Williams & Wilkins Coo, 270, 1941 77. Skow, R. et al., "Hazard Evaluation and Control After a Spill of 40 mg. of Radium," Nucleonics, 11, No. 8, 45, 1953 78. Messler, R. and Widdoes, L., "Evaluating Reactor Hazards From Airborne Fission Products,'" Nucleonics, 12, No. 9, 39, 1954 79. Faarman, A. and Shamos, M., "Effect of Fall-Out from Atomic Blast on Background Counting R te," Nucleonics, 11, No. 6, 80, 1953 80 Lewis, W B "The Accident to the NRX Reactor on December 12, 1952," 80 Lewis, W. Bon "The 1952," Report DR-32, Chalk River, Canada, 1953 81. Hurst, D. G., "The Accident to the NRX Reactor, Part II, Report 6PI14 Chalk River, Canada, October 23, 1953 82. Gilbert, F. WO, "Decontamination of the Canadian Reactor," Chemo Eng. Progress 50, 267-71, May, 1954 83. Russell, Wo Lo, Russell, L. Bo and Kelly, E. Mo, "Radiation Dose Rate and Mutation Frequency," Science 138, 19 December 1958

References Chapter 1 (Contd.) 84. National Acad. of Science "Digest of Findings and Recommendations,"? UoS. Govt.o Print, Office, Wash., D.C., 1956 85. v Cantril, S. T. and Parker, H M., "The Tolerance Dose," MDDC-llO US Atomic Energy Commission, Wash.,, D.C., 1945 86. Carling, E. R. et al, "Radiological Protection - International Commission Recommendations," Nucleonics, 8, No. 1, 31, 1951 87. National -Committee on Radiation Protection "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and[ in Water for Occupational Exposure," National Bureau of Standards Handbook 69, June 5, 1959 88. Cowan, F. P., "Inantitative Summary of Natural Radiation and Naturally Occurring; Isotopes," in Lectures at Inservice "Training Course in Radiological Health" The University of Michigan, Ann Arbor, Mich., 1951 89. Lewis, L. and Coplan, P. E., "The Shoe-Fitting Fluoroscope as a Raddiation Hazard," Calif. Med., 72, 1950 90. "Control of Radiation Hazards in the Atomic Energy Program," U.S. Atomic Energy Commission, Wash., D.C., 1950 91g Lane, Wo et al., "Contamination and Decontamination 'f Laboratory Bench Top Materials,," Nucleonics, 11, No. 8, 49, 1953 92. Ross, D.o., "Cleaning Contaminated Surfaces," Soap Sanit. Chemicals, 27, 1951 93 Brown, R. E. et al.,, "Disposal of Liquid Wastes to the Ground," A/Conf./p 565, United Nations, N.Y., 1956 94. Mawson, C. A., "Waste Disposal Into the Ground," AECL-211, A/Conf./p 1..2; Atomic Energy of Canada Ltd., Chalk River, Ont ''Canada,- United Nations, N.Y. 9, 676, 1956 95. Ginell, W. SO et al., "Ultimate Disposal of Radioactive Wastes," Nucleonics, 12, No. 12, 14, 1954 96. Jensen, J.OHo "Radioactive Waste Disposal in- the Oceana," Nat Bur. Standards, Handbook 58, U.S. Govt. Prints, Office, Wash- D. C.

References Chapter 1 (Contd.) 97.. Seligman, H., "The Discharge of Radioactive Waste Products in the Irish Sea," A/Conf./p 418, United Nations, N.Y., 9, 401, 1956 98. ~Renn, C. E., "Disposal of Radioactive Wastes at Sea," A/Conf.,/p 569; United Nations, N.Y., 9, 718, 1956 99. Taylor, L. S., Paper Presented Before Amer. Nuc. Soc., Wash., D.C., Dec., 1956 100. "The Biological Effects of Atomic Radiation," National Academy of Sciences, U.S. Govt. Print. Office, Wash. D.C., 1956 101. Anonymous, "Nuclear Industry Takes Report on Radiation Effects in Stride," Nucleonics, 14, 1956 102. Braestrup, C. B. and Wyckoff, H. 0., "Radiation Protection," Charles C. Thomas, Springfield, 1958 103.. Recommendations of the.ICRP, Main Commission Report Sept. 9, 1958 Pergamon Press, London, England 104. Recommendations of the ICRP, Report of the -Committee on Permissible Dose for Internal Radiation 1958 revisionPergamon Press, London 1958 105. Maximum Permissible Body Burdens and. Maximum Permissible Concentrations of Radionuclides win Air and Water for Occuupational Exposure Recommendations of the NCRP NaDtiona!llBureau( df.- Standar. ds::Handboo.k 69 106. Addendum to National Bureau of Standards Handbook 59, Permissible Dose from External Sources of Ionizing Radiations, NBS_ April 1958 107. "Rules and Regulations, Title 10-Atomic Energy, Part 20-Standards for Protection Against Radiation," Federal Register, Nov. 17, 1960 108. Davidson, H. 0., "Biological Effects., of Whole-Body Gamma Radiation on Human Beings," The John Hopkins Press, Baltimore, Md., 1957 109. Price.,- W-.,J J "Radiation Detection," McGraw-Hill Book Co., New York,, 1958 110. "Protection against Neutron Radiation up to 30 Million Electron Volts," National Bureau of Standards Handbook 63, November 22, 1!957;

References Chapter 1 (Contd.) 111. Claus, W. D., "Radiation Biology and Medicine," Addison-Wesley Publ. Co., Reading Mass., 1958 112. "Medical Research Council, "The Hazards to Man of Nuclear and Allied Radiations," Her Majesty's Stationery Office, London, June 1956 113. "Effect of Radiation on Human Heredity," World Health Organization, Geneva, 1957 114. Blatz, H., "Radiation Hygiene Handbook,'McGraw Hill Book Co., New York, 1959 115. Neel, J.- V. and Schull, W. J., "The Effect of Exposure to the Atomic Bombs on Pregnancy Termination in Hiroshima and Nagasaki," National Academy of Sciences-National Research Council, Wash., D.C., 1956 116. Novak, J. R., "Radiation Safety Guide, ANL-5574," June 1956 117. National Committee on Radiation Protection, "Permissible Dose from External Sources of Ionizing Radiation," National Bureau of Standards Handbook, 59, September 24, 1954. Ad'dendum, January 8, 1957 118. Glasstone, S., "The Effects of Nuclear Weapons," U.S. Atomic Energy Commission, June 1957

Chapter 2 The Design and Use of Radiation Laboratories The major problem in the design of a radioisotope or radiation laboratory is the attainment of safety, economy, and convenience. These three features, while always interrelated, often seem incompatible~ Consideration of public safety requires that no hazardous materials be released to the environs -by way of the water effluent, the discharged air, or the homeward-bound employees Economy is a consideration of major importance to the industry or institution entering this relatively new field. It is, of course, a relative problem which is quite dependent upon the circumstances. A laboratory that might be relatively expensive as compared to other laboratories on a university campus could at the same time be austere by the standards of latge radiation laboratories. The burden of defining economy, in light of a particular laboratory design, can best be borne by a qualified architect in consultation with a competent radiation expert. Expenditures are justified only if they enhance personnel safety or adequately increase laboratory versatility. The architect should consider critically suggestions relating to such things as (1) strippable coatings on walls and ceiling; (2) stainless steel on all bench tops; (3) thick concrete shield walls around tracer-level laboratories; (4) "special" pipe and valves for low-level waste lines; and a myriad other expenses usually unnecessary. Convenience, with respect to laboratory design, might be defined as "that design which places nearest to the laboratory worker (a) those things which he uses most frequently and (b) those places to which he ventures most often~" Consideration of convenience is important in all instances, 2.1

2.2 including the single-room laboratory as well as the complex multiroom arrangement, Again the problems of a specific design are best referred to the architectconsultant team capable,. on the one hand, of defining architectural practice as it has evolved to date and, on the other hand, of modifying these practices to conform to the needs and restrictions of an isotope or radiation facility. Morgan ( 1) has defined five design variables: the type and level of radioactivity0 the type of investigation; the variety of isotopes and uses, the work volume, and the number of people involved. The applicability of each of these design variables to a specific laboratory is apparent. For instance, the designa of a microcurie-level laboratory for tracer investigations has only one thing in common with the designt of a hot-cell metallurgical laboratory for the invarestigation of the effects of neuwtron irradiation upon metal samples. The comnon bond between these two situations is the presence of "nuclear radiationso In all other respects the two laboratories vary tremendously. The term "hot lab" or "high-level" laboratory is used to describe a laboratory and manipulation facility designed to accomwdate multicurie levels of radioactivity. In its most broad use this term would describe a building with its enclosed shielding facilities together with the required remotely operated tools and handling devices, The function of such a laboratory is to permit the investigator to carry on research,. development, testing, or experimental work, as is normally required, but in a new and different fashion, The investigator works with highly hazardous material in facilities specifically designed to protect him from the damaging effects of the radiation. Safety in hot-laboratory design is most easily attained by adopting a philosophy of containment of all radioactive materials from the time they enter the building as sames until the time they leave the building as concentrated, confined wastes, Thi$ criterion is most satisfactorily approached by providing

High -density concrete 45'-0" ter I. t:-0" _ Im'U De',$sk- I ' Lobb.._6" elev t;;orzz, 4 4 0 ' - 8 ragin ir dWalk in hood 41;'~"1 ~ _ U_ ' TCorridor 60 c24Ionsruteob File conr ee i a aI ecu 2 ppe" wic t I A Mnod 8 e r -lI I II10',_!,,11 ]office ) I",f ic Lobby | | 1-" I.~;.... I;~- ~ II ';o siI.~I Desk I ' 2' - ' ' ' -Irf l ink Book - " 11' Bok ' _~- '~2'-7'1 shelves LAYOUT OF A RADIATION FACILITY Figure 2.1 Floor plan of a proposed concrete-shielded gammairradiation hot ceil -!2 —1....~~.....'~"~"~~I ~j B.. mlnSikr

Figure 2.3 Canyon view in radiometallurgy building at Hanford showing battery of hot cells constructed of cast iron (Courtesy General Electric Co.) Figure 2.4 Ball-socket manipulator in radiation-shield wall (6)

2.3 a pyramid of varying application units0 To il.lustrate: at the top of the pyramid one finds the high-leve '"hot cell"' capable of handling thousands of curies of activity. Since the occasional very highdlevel sample gives rise to numerous intemediate-level samples as cut.off pieces or solutions of a portion of the hot sample, there exists a necessity for a number of the intermediateclevel facilities. The next step down the pyramid finds one faced with confinement requiremnents for lowerlevel nmillicurie quantLties of radioactive materials since the samples$ from the junior caves o shielded gloved4. box facilities will yield low-level samples for further analysis0 In this third step on the pyramid, one would find gloved boxes or hoods with temporary shielding built up to meet the immediate needo The broad base of the pyramid includes facilities for waste handling, decontamination, and sample storage, 2.1 Hotcells The hot cell is a small laboratory space completely isolated from the laboratory worker by heavy shielding walls0 Operations are performed within' the laboratory space by the operator positioned outside the shielding wall and using a remote-type manipulator which operates over or through the wallo The hot cells in existence may be divided into high-level or intermediate-level facilities, depending on their designed capacity for radioactive materials0 They may further be subdivided into open-top or completely closed unitso The open-top units have an upper-limit capacity defined by the "sky shine'. from sources located within the shield enclosure0 This upper limit is in the neigh,borhood of 25 to 50 curies of a gamia-emitting source0 The completely closed hot cell facility may be designed with shielding adequate to handle any contemplated radiation rintensity sourceo Laboratory space should be provided immediately adjacent to the hot cell source it is recommended that these supporting Laboatori~es be located on eigher or both unrestricted wall s a$ ilustrated ina Figure 2lo8 This permits

the installation of plugged holes through the walls for passage of process lines, heating e.lements, control instrumentation, etCo, with the shortest possible runs from the laboratory into the irradiation spaceo The deslgn as pictured in Figure 2.1 incorporates the advantages of an outside loading facility with an under-the-wall water canal. Further, it also includes a laboratory area imdiately adjacent to the ga m-irradiationa room with controlled plugged holes leading through the wall.o The walls of the gamma-irradiation area may be constructed of ordinary or "high-density barytes concrete. If the water pool is designedto pe dd o draining and rePainting, the inner surface may be finished with Amercoat paint. The remainder of the building could be constructed of cement-block wall finished in masonry painet. The floors could be asphalt tile covered, except in the cavs area where they are vinyl or rubber-base cement enamel paLnted (asphalt tile will not withstand continued irradiation), The ports leading thr through the wall from the laboratory to the irradiation room are standard steel pipe collared to elimainate the radiation leakage fromn a straightk-through crack. To use any one of the access holes, the steel pipe filled with concrete would be removed and in its place may be inserted a steel pipe with a helix of copper tubing leading through the pipe, The area sur= rounding the helix would be filled with lead, An alternative arrangement for short-term irradiation studies would be to place the process lines through the tubes and pack lead-shot-filled bean bags into the tuabes around the process lines. Still a third alternative would be to place the copper tubing within the pipe then fill the pipe with a cement=lead mixture (.2)o Figtr 2.2 pictures two htot cels in$tal1ed at the University of Michigan hot laboratory, These hot cells are patterned after the Argonne National Laboratory Metallurgy Cell (-3), Thae cell has walls of high-dens ity concrete

thirty-six inches thick contained with a 3/8"' steel-plate shell. Window openings and ports for the passage of air, gas, water, and service lines were made in the shell before the concrete was poured. Tests on the 30-day aged concrete samples indicate that wall densities of about 220 pounds per cubic foot were attained in this shield. The cells are fitted with lead-glass windows of density comparable to the concrete wall. A set of two Argonne Model 8 Master Slave Manipulators and a thousand-pound pot-lid crane are provided in each of the cells. Inside cell dimensions are 6 by 10 by 12-1/2 feet with a back opening of 6 by 7 feet closed by two fourteen-inch-thilck steel doors. The cell interior (walls, ceiling, and floor) is finished in Amercoat 33. Lighting is provided by overhead fluorescent tubes or soddium-vapor laps mounted around the windows. Roughing filters for exhaust air are located inside the tcave followed by two fiberglas prefilters and a Model A-1000 Cambridge absolute filter located outside and to the back of the cell. The cost (as of 1955) for a single 10,000-eurie hot cell as pictured in Figure 2.2 is listed in Table 2.1. The designer of the hot cell must consider such things as (a) shield structure, (b) ventilation and filtration, (c) manipulator choice, (d) viewing methods, (e) cell lighting, and (f) services to be provided in the cell. The first of these two, that is shield structure and ventilation-filtration tech-o niques, have to do with the safety of the laboratory occupants. The shield may be lead, iron, concrete, dense 3concrete, or any highdensity material which will provide adequate shielding fr6m the radiation sources contemplated within the cell structuree Thicknesses of these materials required will range in the neighlborhood of from 2 to 12 inches of lead, 6 to 15 inches of iron, 20 to 70 inches of ordinary concrete, and 10Q to 40 inches of high-density concrete, It is of first iportancei that the shield wall be of

2.6 uniform thickness and homogeneous construction to minimize the possibility of radiation leakage through a void area within the shield. With this in mindd concrete shields are constructed with care (4). It is itaportant that the interior surfaces of the cell1 be smoth and fissure free to permit easy decontaminationo Table 2.1 Costs* For A 10,000-Curie Hot Cell (5 3 Steelwork, including follow blocks, doors, shell, port plugs, etc, $ 54,200 Barytes concrete (44 yd3 at $120/yd3) 5,280 Windows (3) 30,000 Pot-lid crane 3 400 Manipulators (1 pair Argonne Model 8) 8,000 Manipulator mounting blocks 1,000 Sodium-vapor lau s 1,750 Electrical and sheetmctal 3,500 Door drives and installation 4,100 Painting 800 Contract costs 15_000 $ 127,030 * Based on 1955 prices Multicurie hot-cells constructed of cast iron are used in the radiometallurgy building at Hanford for research with radioactive mnetals (see Figure 2.3)0 The cells afford safe working conditions with highly radioactive alphabeta-anld gaema-emitting materials, wlthile allowing a high degree of flegidlbility ain operation The cells are put together in sectiots and can be dismantled with the aid of an overhead crane. Nuhmerou s plugs in the sides can be remaoved

2.7 to permit installation of special plugs containing lead-glass windows or the insertion of remote-manipulation equipment. Electrical and gaseous services are available inside the cells. The completely closed cell must always be ventilated or exhausted to the point that it is the low-pressure area of the whole building. This insures that all air leakage will be from outside into the cell and prBovides protection to personnel from the possible hazard of airborne contaminationo" In the opentop or canyon type of shield enclosure, air ventilation omst be away from the operator and into the enclosure, All ventilation aLr from the hot cells should be filtered before it is exhausted to the atmosphereo 2.2 2 nipulators The manipulator is a tool or device which translates the operator s movemts from his$ hands to a mchanicaL contrivance remoteLy located, The simplest manipulator would be a pair of tongs which provide the operator with distance protection. The manipulators of greatest interest in hot-cell designt are those which provide the operator with maxsimum dexterity for minimm mental efforto The manipulator type available may be divided into essentially three categories, the ball-socket units, the electrically actuated units, and the pantagraph or mechanical models, Figure 2.4 illustrates a ball-socket manipulatoro It consists of a rod which pushes through a shield ball which in turn rides in a socket mounted in the shielding wallo These manipulators are applicable to wall thicknesses of about 6 inches or lesso Up and down and right to left motion is attained by -moving the ball in its socketo FPront to back motion is attained by sliding the tube through the ball. The tube may be turned to tip the end of the manipulator and it$aemns are grasped by actuating the handle- which opens and closes the tong slave end. (6,7)

2.8 Electrically or hydraulic driven manipulators are available with slave motions in x, y, and z coordinates. The manipulator rides on tracks mounted on the front and rear walls of the hot cello A control console is provided which mounts external to the hot cell and provides control for the x, y, z motions as well as tong gripping, wrist rotation, elbow swing, and shoulder rotation. The main attribute lies in their capacity to lift or move large weights. Electric manipulators may be procured which will handle a 5,00o0lb vertical lift0 The main disadvantage lies in the lack of~ force indication of tong squeeze, The pressure applied by the tong may be indicated to the operator by sounds, lights, or a dial-pointer position, none of which are comparable to the actual hand resistance which one experiences in the mechanical or pantagraph-type manipulator~ Several models of the mechanical pantagraph manipulator have evolved over the past years. The most common unit presently being utilized is the Model 8 Argonne Master-Slave Manipulator pictured in Figure 2020 Figure 2.5 is a sketch of a Model 8 Master-Slave installation showing the mounting conditions as well as the rotary and linear degrees of freedom. (8) 2.3 Viewing techniques There are essentially four viewing techniques available to the hot-cell designer. The simplest of these is a combination of mirrors. A mirror system has the advantage of initial low cost and the disadvantage of revers$ed images at seemingly greater distances, A mirror system of viewing is particularly applicable in open-top cells. A second system of viewing which combines mirrors with an optical-lens system is the periscope. They may be used in open-top cells for tviewinag over the walls with arn assembly in the shape of a large U or they may be used in closed cells by putting a tube through the wallo A movab'le mirror can be

/ I ' /// ~~~~~ Figure 2.5 Argonne Model-8 Manipulator, illustrating the degrees of freedom (connecting linkages pass through the 8-inchdiameter tube. Rollers mounted around the tube permit rotation about the horizontal tube axis). (8) the shield wall (7)

4:4:,"'4 d,, -.,, 4,4 ' 4 4 440,O ' 4 4.. TANK, 4pQ 4,4 4 ~,,?,~' 4,. "~4, d' 4, 4',' 4.' '4 4'',, ':, ", '.,~,, EXPANSION CHAMBER TANK,. 4. " 4,4,, '4', 4~,,,.:..4.. STIRRING PORT 4 4.~' " " CLAMPING BAR-. IJII\ ~, ~~~M"~" GASKETS COATING VENT NONBROWNING GLASS LEAD STEP ~~~~~~~~~~~~~~~~~~~~,.~ LOA IN COMMERCIAL GLASS:09 6~ 0:~9~TANK< SUPPORT Fig~ure 2.7 Isometric sketch showing construction details of a zincbromide window (7),~-.~~~~~~~~W3 9' HOT V~~~~~~;~~~ 'A'~~~ ~ r ~: d SIE..j/~ P. ~'~~t. ~C~':P. LINER ''~ ~~~'~ ': t':'T';,~=r;n 'd~ki~ COMMERCIAL GLA S S '"~~~~~~~'',~,. %::' p.";q'<o~ gp,..o~n ibl~~~AN SPPR '~O~,''~ Fi~~~~~~~~u~~~~~~te~~// Z. smtrcsechsot8cal/l/ndj a:. ~ lc brom~~~~~~~~ide wndo (7 _..,.....% —~ -.,.,~". '~.'.~.. ~;." '/:.'S 4~ '~' dE ~ ~I.~...'-:...; J /,'JJ'/~/,", SID r,'s / Ii/,,!'/,', // b~~~~~~_-. ~:,' Jo=~~~,,~:i"~ ' ~" "~"b' IPZ,,,:,,*..qY HOT I.~:.// 'bY.,: t S~D~ "'"-.- ''!b:~p:4~'a: 6P o ~~~b'b,../ Figure 2~~'6~.8 Ismti kthsoi8cnsrcindtiso e gl~~~~~~~~~~~~ ass shel windo (8)~

2.9 attached on the radiation side, with which the operator may scan the cell horizontallyo Figure 2.6 is a sketch of a periscope for viewing through a shield wall, (7) The most satisfactory viewing of the cell interior is attained by the use of windows through the cell wall. There are any number of window con= figurations available, varying from a small, cylindrical port to a large, square window, There are liquid windows made up of a zinc-bromide solution contained on the viewer's side by a laminated gLass plate and on' the radiation side by a cerium-stablized, nonbrowning, laminated glass plate. These windows have a specific gravity of 2.4, which is comsparable to ordinary concrete, Figure 2.7 illustrates construction details of a zinc-bromide-solution window,(7) Figure 2.8 shows details of a lead-glass window, This window may be const~ucted of lead glass of 3,5 specific gravity, in which case it would provide shielding equivalent to barytes concrete, or it may be made up of plates of glass of specific gravity 6,5, which would approach shielding equivalent to steelo An alternate technique is the construction of a combination leadglass-$zinc-bromide window. The failure of such a window would not leave one completely without shielding, (8) These windows are normally fabricated as a unit contained within a closed tank which is installed in the shield wallo The inside and outside faces are of relatively inexpensive plate glass to protect the lead glass from abrasion and corrosive fumes, The inside plate-glass cover is made up of nonbrowning cerium-stabilized glasso The lead-glass windows are constructed of 4 to 6-inch-thick sheets with a mineral-oil filling between sheets to reduce interfacial reflections and increase light transmittEane, Closed-circugt ate~levsiorn has been usead to a very limited extent in hot-cell viewing t1o dLalteo Th lacLk of applBication appears to arise from two

2.10 causes. There has been little experience in the application of closed-circuit TV in radiation work and, second, there is the problem of attaining binocular vision, with its decided attribute of depth perceptiorn, by available TV systemnso Figure 1.11 shown in Chtapter 1 is a view from Ethe interior of a hot cell. The operator is pouring a radioactive solution into a beaker by use of a manipulator and is observing the operation through a lead-glass window. 204 Hot-cell operational problems If one uses lead-glass windows or remote TV installations, the cell lighting requirements are quite stringent. Precise viewing of delicate operations requires froEm four to five hundred foot can$dles of light delivered in the closed cell, In addition to flourescent or i'ncadescent lights, the cell designer normally provides sodiumovapor lamps capable of providing a single-spectrum light emission which minimiazes the refraction p roblemns in"n herent in multilayer glass windows, One of the more complex operations to perform in a hot cell is the rotating of a crank or faucet. Therefore, all services are arranged on the cell face with controls within easy reach of the cell operator. Services such as air, gas, steam, hot-water, cold-water, vacuum, and drain lines should be' installed on the face of the cell. These services are in turn taken into the hot-cell enclosure through pass-through ports in the shield wall. An alternative arrangement is to put the valve controls external to the cell and the delivery spigots within the cello In addition to the six major variables discussed, consideration most be given to the introduction, removal, and storage of radioactive samples, There are a nuuiber of techniques presently uatilied, the simples$ being the initroduction of the sample cask throudgh doors located behiznd the hot el1, The cask is3 lbroughd t dl n gan the cover removed by means of a crane located

2.11 either between the master$'slave manipulators or on tracks at the ceiling of the hot cello An alternative would be to locate a sa-le entry beneath the floor of the hot cell and hydraulically bring the sample into the hot-cell interior, Then it may be removed from its shield container and placed in position for testing, machining, or otheT studieSo Still a third technique is samle entry into the hot cell from an underwater canal located adjacent to the hot-cell structure. This is a relatively sivele entry technique which provides one with standby storage facilities beneath the water shield. If portions of the radioactive sample are to be removedt from the hot cell it is possible to design samle-remval trays which travel through the ports located in the walls of the hot cello Radiation contminastiQon is a conso tant problem in hot-cell operations., Facilities must be provided for in-Cell cleanup at the comletion of any particular expaeriment These facilities may consist of permanently mounted in-cell washdown hoses and in-cell vacuaum cleaners with attached cyclonepa type separators and absolute filterS, Initial decontamination i$s done with the cell completely closedo When the radiation level is low enough, the cell is entered by the cell operators, attired in protective clothing, shoe covers, respirators, or supply air masks, These people reimove the cell equipment piece by piece to a buffer or a secondary decontamination' zone located external to the hot cell, wlhere equipmlent Is given a final deconltamination or wrapped in protective tape to minimize the possibility of airborne conitaination, and then relegated to storage. Operation of a hot cell requires continuous mechbanical maintenancel togetlter with intriccte design a d fabrication of mechanisal devices capable of being reotely operated within the cello This designs fabrication, and maintenance work is best perfor sd by shiop facilBities located in t he iediate

2.12 vicinity of the hot cell, The operating personnel of the hot lab must be provided with lockers, clothing-change rooms, and office space. 2.5 Junior caves and shielded glove boxes The intarmediate-level radiation sources of L to 100 curies are handled remotely in junior cave assemblies or lead-shielded gloved boxes Figure 2.9 is a photograph of one of the junior caves at Hanfordo It consists essentially of a chemical laboratory bench, complete with water, gas, and other senrvices surrounded by a thick steel shield to protect the operator from radiation while working with radioactive materials. The operator can see what takes place through a thick window consisting of laminated plate glass or lead glasso The window is thick enough to provide the same protection from radiation as several inches of steel. The work is performed by means of remote handling devices that consist of rods placed in ball-and-socket joints in the front of the cave. These rods are equipped with pinchers that can be opened and closed to pick up glassware and other equipment and can be moved laterally and vertically in the ball joints or back and forth through sleeves in the joints. Figure 2.10 is a photograph of two gloved-box units, The unit on the left is a 214/2 inlch lead-shielded assembly for remote chemistry worko There are two ball-socket manipulators through the shield wall. The shield assembly is mounted on casters and rolls into position around a standard box such as the unit on the right in the photograph. Services for the gloved boxes are provided from the enclosed pipe chase pictured behind the boxes. Services provided are gas, water, air, vacuum, waste-water drain, and filtered exhaust-air outlet, The shielded box is a small (about 9 cubic feet) self-contained laboratory whicoh can be used for remote chemical, metallurgical, or biological

Figure 2.9 Junior cave used at Hanford (Courtesy of General Electric Co.) Figure 2.10 On the left, shielded-box assembly with ball-socket manipulators. On the right, standard gloved-box unit

Figure 2.11 Interior view of a shielded-box unit equipped for remote radiochemistry work (Courtesy Kewaunee Mfg. Co., Adrian, Michigan) Figure 2.12 Photograph of a radiochemistry laboratory, showing the - hood installations and gloved-box manifold assembly on the right (5)

2.13 investigations~ Figure 211 is a photograph of the interior of a shielded box with remete-operated equipment for radiochemical separations0 Table 2.2 is a cost summary for a basic shielded-box unit. Table 2 2 AoProximate Cost For A 1 to 10-Curie Shielded Box* te pproximate C:ost Basic glove box $ 59500 Mobile stand for box 50o00 Lead=shield assembly, 2-1/2 inches 2,000.00 Two ball-socket tong manipulators 714.00 Fluorescent box light 75.00 Power panel assembly 52000 High-efficiency exhaust filter 46o00 Lead-glass window 1,400.00 Miscellaneous small parts 200,00 (trays, flexible duct, etc.) _ $ 5,132.00. -- *Courtesy Kewaunee Mfg. Co., Adrian, Mich. One of the attributes of gloved boxes is the small-volume exhaust-air requirement0 lost box installations are operated at 5 to 15 cubic feet per minute of exhaust, The box units should be used as open-front hoods for low-level work only. since the airflow velocity into the box with the glove panel removed will be from 2 to 10 linear feet per minute0 This is not suf-7 ficient airflow to guaranatee transport of fumes or particulateo Figure 2o 12 is a photograph of a chemistry laboratorry imstallation, The

2.14 hoods pictured are equipped with a bypass airflow arrangement which provides a constant airflow face velocity of 50 linear feet per minute independent of the area of the face opening. These hoods have rounded.corers to eliminate turbulent airflowo Air filtration is provided by filter boxes mounted above each hood. The filter boxes contain fiberglas prefilters followed by Model A-1OO Cambridge absolute filters. Figure 2o13 is a photograph of a walk-in hood installation, This hood is used to house large organic-syntheses assemblies for the preparation of carbon-14 labeled compounds. The exhaust-air filter boxes are mounted above the hood~ The two manometers visible at each corner of the hood indicate pressure drop through the filter box and provide a constant check on the extent of filter dirt load. 2.6 The low-level or radioisotope laboratory The simplest design for a radioisotope laboratory uses one room and contains a hood (or glove box), a sink, bench space, and storage room. A plan for a typical design for such a laboratory is given in Figure 2.14 and 2.15. The 300 square feet of floor space indicated provides areas for experimentation, sample assaying, and paper work. The desk section is lower than the bench section so the investigator may work comfortably when seated. A source storage location may be provided by building a temporary lead-brick shield assembly either under the glove box or in the storage closet. This one room laboratory contains essential working facilities and is suitable for low-level tracer research work. The restricted quarters in the one-room laboratory do, however, present some operational difficulties. First, there is the possibility of contaminating the coulting equipment and desk section ote one e side of the room with readioactive materials from the experimental side of the ro8o The experietnter

-iii;-i~ ~~ ~............?.lii~ i-iii~iiii-'ii~a~i~ii:::i~ i i: Figure 2.13 Photograph of a "walk-in" hood capable of housing assemblies 6-1/2 feet high. A standard chemistry hood-bench assembly is visible in the background (5) ~~~[ I~~20 0 8' 7" 3 0 7' 6" 0 in 0 2 4 Figure 2.14 Plan view of typical one-room radioisotope tracer laboratory (Courtesy of Kewaunee Mfg. Co. Adrian, Mich.)

Figure 2.15 Isometric view of typical one-room radioisotope tracer laboratory (Courtesy of Kewaunee Mfg. Co., Adrian, Mich.) 'II Figure 2.16 Isometric and plan view for a typical three-room radioisotope tracer laboratory (Courtesy of Kewaunee Mfg. Co., Adrian, Michigan)

2.15 must exerclise extreme care in order to prevent the spread of traces of contaminationo Another difficulty is the presence within the room of the radioactive source marial which may interfere with the counting of low-level radioactive samples. If relatively large quantities of radioactive source materials are ordered and stored in the lab, the unused portions must be shielded to keep the background radiation at a minimumo A question is sometimes raised regarding the difference between this radioisotope-laboratory design and other typical chemical-laboratory designs. Actually, there i Wno significant difference and existing laboratory facilities are entirely adequate for most low-level tracer eaxperiments that can be performed with radioactive materia3ls, Only the techniques of the investigator amust be changedo The need for strippable coatings, stainless steel, and special finishes has frequently been overemphasizedo Since the problem of wall and ceiling contamination in the low-level tracer laboratory is practically nonexistent~ a standard wall finish such as painta on plaster or cement block gives'a durable, easy-to-maintain finish. The floors may be asphalt tile, mastopave, linoleum, or any other comparable sealant covering that is comfortable to stand on0 If two rooms are available for the laboratorys the logical arrangement would place the office operations and counting equipment in the second room and leave the first room exclusively for the experimental work0 If three rooms, can be made available, the counting equipment could be further isolated from the laboratory by placing the office between the laboratory and co unting room. An alternative arrangement would be to establish a low-level laboratory, a higher-level laboratory, and a coamabination counting room and officeo The decision as to which layout would be mst.satisfactory is too a large degree dependenxt on the type and level of radioactive material whsich will be aeeded

2.16 to carry out the desired investigations. An isometric view and floor plan for a typical three-room radioisotope laboratory are shown in Figure 2.16. One room is used for all experimental manipulations. The middle room is an office and the third is a counting room, Such a laboratory provides ample work space for two research people and two technicianso The estimated cost for furnishing each of the three laboratories is $6,000 for the one-room tracer laboratory, $8,000 for the two-room combination, anid $10,000 to $12,000 for the three-room combinationo These estimates include the furniture indicated in the figures but do not ancluade the:cost for,plumbing, lighting, wall finishing, or an air-exhaust system.o -Table 2.3 summarizes costs for the initial establishment of a laboratory. Table 2.4 gives an estimted budget for equipment for a one-room laboratory using a single radioisotope. Table 2.5 lists appropriate instrumentation (and approximate prices) for a versatile three-room radioisotope laboratory. Estimated annual operating expenses for the small laboratory are given in Table 2. 6 Additional informaortion on te use and deign of radioisotope and radiation laboratories is given in references 37 to 117.

2.17 Table 2.3 Radioisotope Laboratory Equipment and Alterations Cost (9 *' One-.Room Two -Room Three-Room Lab Lab Lab Minimum $ 3,700 $ 4,300 $ 6,900 Typical 6,400 7,500 12,900 Maximum 9,100 11,800 21,300 Table 2.4 Small Radioisotope-Laboratory Equipment Costs (10) ** Safety Instruments $. 715 Counting Instruments 1,800 Shield Equipment 2,020 General Laboratory Equipment 1,500 $ 6,035 * Based on 1952 prices ** Based on 1956 prices

2.18 Table 2.5 Equipment For The Versatile Three-Room Radioisotope Laboratory Item Number Total Cost Scalers 2 $ 1,600.00 Vertical lead shield 1 285.00 Geiger tubes 2 100.00 Q-gas tank and regulator 1 93.00 Flow-gas counter 1 395.00 Scintillation-well counter 1 970.00 Geiger-type survey instrument 1 300.00 Ion-chamber-type survey instrument 1 300.00 Count-rate meter 500.00 Recorder 1 400 00 Pocket ion chambers 12 120.00 Charger-reader for pocket chambers 1 225.00 Film-badge holders 12 12.00 Counting calibration standards (set) 1 80.00 Laboratory cart 1 50.00 Lead shield bricks 20 200.00 General laboratory equipment (centrifuge, refrigerator, glassware, hot plates, balance, pH meter, etc.) 3000o00 $ 8,630.00.

2.19 Table 2 6 Estimated Annual Laboratory Operating Expenses Of The Small Radioisotope Laboratory (10)* Salaries, Administration, and Service $ l3,000.00 Laboratory Supplies 1,000.00 Consulting Services and Travel 1,500.00 Building Space and Services 1,500.00 Depreciation and Maintenance 1 000.00 Analysis and Testing 200.00 Miscellaneous 250.00 $ 18,450.00 * Based on 1956 prices 2,7 Common hazard parameters All radiation and tracer laboratory design concepts must include consideration of the two hazard parameterso That is, the external radiation hazard and the internal radiation hazard. The former is dealt with by the erection of proper shield walls, whether they be temporary or fixed as in the case of the gamma-irradiation facility. The latter, the internal radiation hazard, is dealt with by providing glove-box facilities, hood facilities, radioactive-waste drain systems, exhaust-air filter systems, and general control of contamination. A first step toward general control of contamination is taken when one lays out the laboratory, grouping the higher-level radiation laboratories or high-risk facilities away from those areas of long-term occupancy or favored intake such as office space and lunch roomso 2.8 Exhaust-air control In general, one tries to lay out the laboratory facility in such a

2.20 manner that minimaum traffic is directed into the high-level areas of the labo In turn, the airflow throughout the total facility is directed from areas of occupancy to areas of high-level operations as illustrated in Figure 2l17o There are two alternative techniques for air filtration from radioactive facilities. One is to place filters immediately adjacent to the point of contaminationo For example, some facilities have filter boxes located directly above the hoods, as illustrated in Figure 2.18. Located in this manner, a high contamination incident within the hood contaminates only a single filter unit. The second alternative method is to place a single large filter bank to service all the hood facilities0 This latter technique results in total filter-bank contamination in the event of a single hood facility contamination incident~ The problems inherent in the design of an exhaust system for handling radioactive off gases or dusts are dealt with quite extensifvely in a publication by the Building Research Advisory Board (11). 2.9 Monitoring with instruments Electronic radiation-monitoring devices carried or strategically placed in the vicinity of an experiment invrolving radioisotopes can be kept in continuous operation throughout the experiment. These devices can indicate the instant a predetermined "'safe" level of radioactivity is exceeded. Typical instruments used are: (1) meters for recording fast neutrons, known as "Neuts"; (2) Geiger counters for low-level monitoring; (3) portable ionizing chambers called "Cutie-Pies " (4) beta-gamma meters with probes; (5) meters to detect the presence of alpha-emitter substances, termed "Poppies"' and (6) "Juno" meters, widely-used general-purpose meters that can detect alphasparticle radiation as well as beta and gaamna, Fire 2.19 shows a fixed ionization chamber which constantly records the radiation level behind a massive concrete wall. It: indicates precautions

COLD LABS Is i Montor Rrn stair I ing Rm stair l'Ir ~ i A1L Ft no.4 o: CORRIDOR - CONDITIONED AIR 'I-Il rT ATIC -1F INST. ONTAMINATIO ROOM ~r r tr Fr o t 1n, SUPPLEMENTARY HOT CORRIDOR. CONDITIONED AIR VAULT SAMPLER HOT LABS Figure 2.17 Floor plan of a radiochemical laboratory building with radiation zoning and airflow planning (6) Figure 2.18 Double hood installation with exhaust-air filtration boxes above. (Courtesy Kewaunee Mf. Co., Adrian, Michigan)

Figure 2.19 A fixed ionization chamber which measures radiation level at point inside shielding wall (Courtesy General Electric Co.) contamination before leaving building at Hanford Plant (Courtesy General Electric Co.):s~~~~~~~~~~~~~~~~~~....

2.21 required for personnel entering such a zone~ Permanent instruments can provide a continuous record of radioactivity levels at a remote location, or a warning signal to enable disaster procedures to be set into instant operation to minimize exposures (12g. Figure 2Q20 shows a final radiological check of workers' hands, feet, and clothing as they leave the laboratory. This procedure is good protection for the worker as well as a positive prevention of the spread of radioactive material from the laboratory. Handmand-foot counters are located adjacent to areas of possible contaminationl These enable employees, leaving a building after removing protective clothing (or after walking down a hallway also used by people who have just removed protective clothing) to stand on one pair of instruments and place their hands next to another set of instruments that measure and record the amount of radiation being given off by material on the skin or shoes. This is a final check to determine if the protective equipment and pro" cedures have performed their functions properly and to assure that the employees are free of contaminationo That such measures can be very successful is shown by the record at Hanford where no employee has ever been seriously or permanently injured by radiation or by radioactive material. 2Q10 Personnel film badges In the routine operation of any radiation laboratory or area of significant radiation level, records must be kept of the radiation dosages received by all personnel having access to the radiation areaso This is usually accomplished by the use of film badges such as that shown in Figure 2.9. For the personnel this is essential to prevent excessive and harmful exposure to radiation as a result of continual working about an area of high radiation level0 Also, such records provide protection from legal action by eigher employees or visitors

2.22 who unjustly claim over-exposure to radiation. The same bsadge is worn daily by laboratory employees and the film packet is changed periodically, usually once per week or twice per month or immediately after a radiation incident of suspected over-dosage. The exposed film packets are returned to the supplier where they are developed and returned with a report of the accumulative dosage given each badge. Nuclear radiations and x-radiation darken the x-ray type films that are used in the badges in a manner similar to the darkening of photographic film as a result of the exposure of the film to light. Both the x-ray and ordinar) photographic films consist of a light-sensitive emulsion layer deposited on a cellulose acetate or glass base. The light-sensitive emulsion layer consists of a colloidal dispersion of crystals of a silver halide (or a mixture of silver halides) in gelatin. All emulsions show a great variation in sensitivity to the energy of the activating x- or gamna-radiation. In practice this is taken into account by covering the film with filters that reduce the sensitivity of the film, and flatten its response to radiation of various energies. For a given film the degree of darkening is dependent upon the properties of the film and the type, duration and intensity of the radiation that reaches the filpn, To cover a wide radiation dosage range, two or more films are usually usedO If two films are used in the packet, one of the films is usually highly sensitive and may cover the range of radiation dosages from 0 to 2 roentgens, The other film usually is less sensitive and may cover the dosage range up to 30 roentgens (300 times the maximum tolerance for one week), The two films are wrapped together in thin opaque paper to keep out light which would expose the films~ In one type of badge a portion of the film is located at a "window" and a portion behind a thin cadmium shieldo In the case

2.23 of exposure to mixed radiation (beta and gamma) the beta particles are stopped by the cadmium shield but the penetrating gamma photons pass on through and expose the film. Beta particles readily penetrate the film packet at the window and expose the film but do not penetrate cadmium shield. In another type of badge:a stepped copper wedge is used irn place of the cadmium. This type of badge is particularly useful in determining dosages resulting from exposure to x-radiation. In case of x-radiation the film density varies greatly with the kilovoltage used in the machine. The use of copper steps permits the comparison of the film density under each step with standards exposed to known dosages of xmradiation produced by machines operated at known kilovoltages. If there is a possibility of neutron exposure a third film which is sensitive to proton radiation is placed behind the "beta-gauia" films. With such a film badge fast neutrons pass through both the window and the cadmium barrier, but slow neutrons are readily captured by the cadmium and, therefore, pass only through the window.,The fast neutrons interact with hydrogen atoms in the film to produce recoil protons that expose the film. The slow neutrons also produce protons from the N14 (n,p) C14reaction giving an additional proton exposure. The protons produced by both processes leave "tracks" on" the third film. The dosage for fast and slow neutrons can be estimated by counting the tracks after development of the film. This involves microscopic analysis and use of the phase microscopy principle, One commercial laboratory uses Eastman Type NTA film with an emulsion thickness of 25 to 30 microns, calibrates the film with polonium-beryllium neutron sources, and analyses the film at 860X. The beta-gamna dosages are obtained by comparison of the film density at the window and behind the cadmium with that of calibrated control films

2.24 exposed to known dosages of radiation. To avoid errors from variables in developing, control films are developed and fixed with the test films. 2,11 Use of special clothing Radioactive material may enter the body by three main routes: inhalation, ingestion, injectiono Air-borne particles or radioactive gases, such as C1402, may be inhaled and incorporated into the tissues. Radioactive materials unconsciously or absent-mindedly transferred to food, cigarettes, or pencils may find its way into the body through the mouth. The accidental puncturing of protective clothing and the skin during work may inject radioactivity directly into the blood stream0 Accidents, or "spills," in the laboratory occasionally release radioactive materials into the air and on to laboratory surfaces(13)o Ventilating systems must not exhaust radioactive gases or dust inato the atmosphere and sewage effluents must not release radioactive materlals into lakes and rieers. Laboratory workers must not carry the radioactivity from the laboratory on their bodies or clothes to "clean"' areaso Figure 2.21 shows a worker returning from an operation in a highly contaminated area at the Hanford Atomic Products Operationo The special suit provides adequate protection from contamination of the skin surface as well as inhalation exposureO Before removing the outer layer of protective clothing, a radiation monitoring unit worker (health physicist) checks the clothing with an instrumant. This worker is wearing full protective clothing including two pairs of gloves. The space between the inner gloves and the sleeve of the coveralls is sealed with masking tape to prevent air-borne contamination from getting inside, The hood-type head cover is tucked inside the outer coverall baut outside the inner coverall, Mask is fresh-air type that furnishes air from outside the contaminated work area.

Figure 2.21 Worker with special suit for protection in radiocontaminated areas being checked for contamination. (Courtesy General Electric Co.) Figure 2.22 Routine check of floor for possible contamination (Courtesy Atomic Energy Research Establishment, Harwell, England)

Figure 2.23 Radioactive liquid-waste tank farm of 10,000-gallon storage capacity. Michigan Memorial-Phoenix Laboratory, Ann Arbor, Michigan

2.25 The special clothing worn by the worker shown in Figure 2.21 (and also by workers shown previously in Figures 1.2, 1.3, and 1,9) is of primary importance when working in areas in which the atmosphere is possibly contaminated by alpha-particle emitters such as plutonium. If these same workers were using encapsulated alpha-particle sources or working with gamma radiation in uncontaminated areas, conventional laboratory clothing with suitable radiation shields and survey instruments would be sufficient protection, (See Figures 2.3, 2.9, and 2.10) Alpha-particle radiation is completely absorbed in a small volume of tissue leading to severe localized tissue damage, sarcoma, and related injuries. If an equal amount of energy in the form of beta radiation is absorbed in the body, it is distributed throughout a much larger tissue volume. Tissue damage in this case is more diffuse but still localized. Internal gamma radiation would penetrate still further and, in fact, be expected to produce no local injury, only gross, total body radiation damage. Thus, while gamma radiat ion is more damaging than alpha radiation when the exposure is from an external source, the reverse is true if the source is internal. 2.12 Decontamination procedures Decontamination may be required for-protection against excessive radio ation levels or radioactive material entering the body. The risk of radiation injury due to ingested, inhaled or injected radioactive materials can be minimized by careful planning, continuous radiological monitoring and good housekeeping procedures, The initial design of an experiment or process must take into account the degree of contamination hazard involvedo For instance, in general, all work with alpha-particle emitters must be perform9ed in closed systems, such as gloved fume-hoods or glove boxes, Good experimental design can minimize the

2.26 possibility of an incident which might spread contamination, as well as include safe procedures in the event of such an incideant The procedures involved in keepinag a laboratory radiologically "cleaw are no differentt ina principle froml those used in good housekeeping procedures, No dust or dirt is allowed to collect. Waste is disposed of before large amounts accumulate. Smooth-finish surfaces, such as stainless steel, are used when possible. Otherwise, cheap, easily removable surfacing material is used, Porous surfaces such as unglazed ceramic tile, wood, mortar or concrete are avoided or are coated or covered with nonporous materials. This type of laboratory operation coupled with suitable monitoring procedures can minimize the risk of workers carrying radioactivity away from the laboratory. However, in spite of all precautions, human error, mechanical failure, and accidents could be responsible for serious instances of radio-contamination. In such cases, protection consists primarily of detailed surveys to delineate the dangerously contaminated areas and of keeping away from these areas until decontamination is completeo Decontamination, the removal of radioactive material from an area, is much the same in principle as the removal of any other contaminant (14)o, Soap and water, and abrasives when necessary, are satisfactory decontaminating agents (15)9 This process differs from other types of cleaning, however, in two respects. The contaminant is not normally detectable to the sight, smell or touch. Frequent surveys with a Geiger tube or ionization-chamber instrument during the decontamination procedure are necessary to evaluate the efficiency of the procedure. Figure 2.22 shows a routine check for possible contamination being carried out at Harwe1, England. The material removed, including detergents and wash liquids remaining in mops and clothes, is itself a radiation hazard and cannot be disposed of by

2.27 ordinary means. These materials, depending on the degree of hazard involved and other factors, must be diluted before disposal, stored for decay, or shipped to a site of permanent radioactive storage, as must any radioactive waste material. Table 2.7 outlines factors to be considered in area decontamination. Additional information on this subject is given in references 16 to 21. Table 2.7 Outline of Factors to be Considered in Area Decontamination (*) I. PRE-INCIDENT PLAN4NING: A. Proper Facilities 1. Laboratory design 2. Experiment design B. Isotope Choice 1. Properties required 2. Example: Sr89 vs Sr90 C. Experience available D. Assistance needed E. Equipment required to handle contamination incident 1o Low-level liquid (a) Absorbent paper (b) Trays (c) Mop and bucket (d) Shoe covers 2. High-level liquid (a) Isolation equipment (signs, barriers, etc.) * Prirvate communication from A. H. Emmons

2.28 (b) Flush hose (in cell) (c) Mop and bucket and detergents (d) Waste storage container (e) Shoe covers (f) Protective clothing (g) Shower facility 3, Low-level dusts (a) Respirators (b) Vacuum cleaner (c) Shoe covers (d) Caps 4, High-level dusts (a) Supply air mask (b) Vacuum cleaner (c) Shoe covers (d) Protective clothing (caps, coveralls, etc.) (e) Shower facility (f) Waste storage facility (g) Isolation equipment (signs, barriers, etc.) II. ACUTE PHASE: (Incident has Occurred) A. Evacuate (but don't go home') 1. Not necessary with low level liquid 2. Call for Health Physicist assistance 3. Think' B. Assemble equipment to begin clean-up C. Assign specific duties D. Be prepared to rotate persomnnel (if necessary)

2.29 E, Obtain an air-sample (if possible) III. RECOVERY PHASE: A. Instrumentation-Take correct instrumentation capable of detecting the contamination into the contaminated area B. Plan and execute decontamination procedures C. Evaluate the efficiency of each procedure D. Rotate personnel E, Keep calm and use good sense! IV. TECHNIQUES: A. General 1. Leave the area (or object) vacant and allow natural decay to reduce the levels. 2. Decontaminate B. Types of Decontamination,. Rough decontamination (limited use, or occupation of area) 2. Detailed decontamination (restore to original condition) C. Processes of Decontamination 1. Wet contamination on smooth surfaces - Flush with water and detergents followed by scrubbing and flushing. Steam under pressure is very helpful. Treatment is more difficult if the contamination has been allowed to dry. 2. Contamination in form of dust - Use vacuum cleaning. If further treatment is necessary, brush and vacuum again. 3. Contamination of greasy surfaces - Remove greasy material with dry-cleaning solventso If necessary, follow by scouring with water, soaps, and detergentso 4. Deeply absorbed contamination - Removal of surface, at the same

2.30 time preventing further penetration of contaminant (e.g. removal of top few inches of earth, caustic paint removers, abrasion, acids or special chemical compounds). 5. Firmly-held contamination - Covering with suitable thickness of sealing material, or disposal of contaminated objects or clothing by deep burial in the earth or at sea in weighted, sealed container. Decontamination may have to be carried'out in, several stages, and a combination of methods may be necessary. Do Facts about Contamination 1. It will not be uniform 2. Smooth surfaces less susceptible than rough 3. Cracks or crevices collect it 4, Movement of men and equipment will spread it 5. The more porous the material, the tougher the job. 6. No process will neutralize it E. Details regarding rough decontamination 1. Speed is main consideration 2. Simplicity is usually forced 3. Most practical technique is water wa Add soap or detergent if available. Use hot water if availableO 4. Land areas (a) Bulldozer (b) Graders (c) Wet it down first 5. Best protection - distance' 6. Points to consider: (a) Method mast work quickly

2.31 (b) Method must be suitable for material (c) Should not require large quantities of special or dangerous chemicals (d) Should make use of available equipment, services, and material F. Procedures for detailed decontamination 1. Summary of methods applicable to surface decontamination: (a) Vacuum cleaning (b) Water (c) Steam (d) Detergentb (e) Complexing agents (f) Inorganic acids (g) Organic solvents (h) Caustics (i) Abrasioli (j) Flame cleaning (k) Remove outer layer of material with contamination 2. Disposal techniques (a) Burial on land (b) Entombment (c) Burial at sea G. Personnel decontamination 1. Soap and water scrub (no water, then wipe) 2. Remove contaminated clothes 3. Scrub 4 Check

2.32 5. Scrub 6. Stay out of clean areas unless you are cleang 2.13 Radioactive wastes For the small tracer or isotope laboratory, the problems of radioactivewaste handling are relatively minoro Short-lived liquid wastes of low-level concentrations may be dumped directly into the sewage lines, provided concentrations and quantities are kept below the limits established by the National Committee on Radiation Protection and local codes. Higher concentrations of short-lived wastes may be retained for decay and then disposed of with adequate dilution. The short-lived solid wastes may be stored. Those things of value may be recovered by decontamination procedures. Those things of no value, such as paper towels, rubber gloves, used or broken laboratory equipment, may be stored for decay and eventually disposed of in a moral fashion. A suitable solid-waste handling technique for a tracer laboratory, using a short-lived tracer, might be illustrated by consideration of an average problem. Consider a laboratory which receives 55 millicuries of P32 every 14 days. If all this P32 is disposed of into a radioactive-waste can, 55 millicuries, less that lost by decay, will be added to this can every two weekso Eventually an equilibrium concentration will be reached of 109 millicuries of P32 in the can. This concentration of activity would be attained after approximately 70 days. If this filled can is now replaced with an empty one and the filled can is placed in storage for 120 days, there would be 0.27 millicurie of P32 activity remaining in the can, At that time, the filled can could be removed from storage and the contents burned in an incinerator, with no problem. These numbers, of course, are quite unrealistic ill that it is highly improbable that one would discard the total P32 shipment every fourteen days. A more realistic value would be three to six per cent of the shipment going into the

2.33 active solid-waste containers. Then after 128 days decay, one could expect a concentration of eight to 16 microcuries in the filled waste can. For the small tracer laboratory, the easiest approach to the liquid-waste problem is to confine all suspected liquid wastes to one-gallon waste jugs placed conveniently in the corners of the hoods. After a number of jugs of liquid wastes are accumulated, a sample is taken from each and assayed to determine the activity levels. If the jugs display long-lived radioactive contamination, the user is faced with two alternatives. Re may process the waste into concrete by using the liquid waste as the water component of a concrete mix, then designate a controlled area as storage for the final blocks. A second alternative is to procure the services of a company handling radioactive waste*, or make arrangements to ship the wastes to a suitable government-operated burial facility, For the larger laboratory specifically designed for work with radioactive materials, it is often most convenient to connect the drain located in the floor of the hood directly into a five-gallon polyethylene carboy. The radioactive liquid wastes are collected in this five-gallon carboy. In the event of a spill within the hood, the -spill as well as the contaminated wash water drains directly into the carboy. If this approach is taken, it is important that one locate any drains connected with the sanitary sewer system above the floor of the hood in such a manner that nonactive run-off water must be actually placed in the sewer line, * Radioactive-waste disposal service Several companies providing are: Crossroads Marine Disposal Corp. Radiological Service Coo 26T Warf 92-15 172nd Sto Boston, Mass. Jamaica 33, N, Y, Reed-Curtis Nuclear Industries, Inc. 307 Culver Blvdo Plaza del Rey, Calif,

In the large, integrated, ltllaboratory facility it is advantageous to lead all the drains into a retention tank with a liquid-level indicator mounted within the laboratory facility. In this arrangement the normal laboratory wastes are allowed to accumulate within the retention tank. As the tank fills the wastes are monitored to determine the activity level. If the activity level is below maximum permissible dump level, the tank is drained or pumped into the sanitary sewer system, If the wastes in the tank are above permis'. sible dump levels, they are pumped into storage drums for concentration or long-term storageo Figure 2.23 pictures a retention-tank "farm0' for radioactive wasteso Often this latter system of retention tanks is ope rated together with a norma1 sanitary sewer system, where the sanitary sewer drains (or the hot drains) are painted a distinctive color to identify one from the other. This approach minimizes the total volume of possible contaminated o'St$h o wastes. For a very large laboratory facility, particularly one operated itn conjunction with a reactor, it is essential that space be reserved for the installation of som type of concentrating equipment capable of handling low- to medium-level radioactive wastes. Various approaches have been made to this problem. In general, if the wastes are low in total salt content,,one may use ion-exchange equipment. If-the wastes are of high salt content, one is forced to use flocculation, evaporation, filtration, or some similiar concentrating technique (22) o Plans and procedures for such "1ultimate" disposal from the smeller laboratories usually involve either burial in the ground or dumping into the sea, The release of radioatctive wastes directly into the ground relies on the filtration, absorption, anad ion-exchange properties of the soil (23,24) to fix the radioactivity within a limited, controlled volume and thus prevelnt

2.35 contamination of local bodies of water or water supplieso When geological, meteorological or ecological considerations do not justify the assumption that the activity will be retained, artificial absorption on clay (25) or "permanent" storage in steel and concrete tanks may be undertaken. Because so little is known about the diffusion and mixing properties of the oceans, and the biological processes by which isotopes may be concentrated by marine organisms (26), the disposal of large quantities of waste into the sea for dilution may not be feasible. No maximum permissible concentrations of radioactive material in sea water have been set because of these uncertainties. Permanent packaging and storage in areas remote from fishing areas and having geological characteristics favorable -to long-term undisturbed storage seems desirable at this time (27). Ocean deeps and deep-sea ooze may not be as promising storage areas as many believe, however, owing to oceanographic disturbances (29). Additional information on the problem of disposal of radioactive wastes is given in references 116 to 124. 2.14 Example design of a multipurpose hot-lab Figure 2.24 is the first floor plan of the Michigan Memorial-Phoenix Laboratory located at the University of Michigan. This floor plan will be used to illustrate some safety features sad design concepts which have been incorporated into this building. This laboratory is designed for use in a wide range of research studies and for work with radioisotopes having activities ranging from microcuries to kilocuries of gamma activity. Figures 2.10, 2,11, 2012,92139 2.18 and 2.22 show views of some of the installations in this laboratory. Entrance to the hot-labo>ratory section is by way of a short side-corridor opening into the restricted-access corridor. This short corridor passes by the locker-change room and the health-physics officeo The short corridor is used as a pick-up point for film badges and pocket chambers and houses a hand

2.36 and-foot counter for personnel monitoring0 As one enters the restricted-access corridor and progresses toward the far end the installations become ever-increasingly "hot." Behind the two hot cells is a large isolation and decontamination area for the cleanup and storage of contaminated equipment arising from hot-cell operations. Beneath the floor of the isolation-decontamination area there are three large storage tanks, as pictured in Figure 2.24. These tanks hold the radioactive waste water from the chemistry laboratory, hot cells, and all hood installations throughout the building. The laboratory is equipped with a double drainage system. All floor drains, hood floor drains, sinks in the decontamination area, and certain sinks in the low-level area drain into the active waste-storage tanks, The waste-storage tanks are designed to handle up to millicurie levels of radioactive liquid wastes, Wastes of levels higher than this are placed in collection bottles within the hot cells and chemistry hoods. These bottles are in turn collected and placed in storage, The radioactive liquid wastes are decontaminated by sandbed filtration and mixed-bed ion-exchange resin colunls. A second sanitary drain system in the hot-lab area is arranged in such a manner that accidental entry of active wastes can be made with difficulty. This installation of the sanitary system minimizes the amojnt of water that must be treated in the decontamination facilities. -The rear wall of the decontamination area is below grade-level with a storage facility for radioactive materials. There are 31 sample storage holes and five lead-shielded sample storage drawers built into this below-grade wall. There is a fifteen-ton hydraulic lift, capable of rising above grade to trck-bed height, provided in the decontamination-area rear, It is possible to place a fork lift on the elevator, rise up to truck-bed heights, lift off a large shipping container, and amove it down to the hot cells. (5)

HYD SRAUM A M LIFT ESCAPE DISPOSAL C' DECONTAMINATION ELECTRICAL ROOM SA H SPACCELERATOR ROOM HIGH GAMMA MECHANICAL EOUIPMENT ROOM AREA CAE I G AVE OPEMEALCKRRO ROOM JANITOR'S MEMORIA LO R -nR FIRST FO PLAN CLOSET SHAFT 2.24 Mhiga Mo Peix L a h br SHAFT UTILIT UTILITY SHA ITY SHAFT ILI tOSET LIT ~~~~~~~~~~SHAFT PIPE TUNNEL 1 -— CORR IDOR ---- I~T WL YO-) PEDESTRIAN TUNNELRACO BIDIG womEN's LOCKERRO COUNTING UNASSIGNED SPACE NI OT SPECIMEN SHIELDED BOX IIHIGH LEVEL CHEMISTRY HEALTH ROOMROO MEN'S LOOKER ROOM IPHYSICS ROOMROMOM PHOENIX MEMORIAL LABORATORY — FIRST FLOOR PLAN Figure 2*24 Michigan Memorial-Phoenix Laboratory hot-lab floor plan

The second floor of the Michigan Memorial'Phoenix Laboratory is at grade level and houses the administrative and nonradioactive functions of the laboratory., Office spaces, conference room, library, machine shop, and an electronic shop are located on the second-floor area. Accidents will happen in the best designed and staffed laboratories. Preparation in advance will reduce personnel injury, property damage, and assist in the maintenance of operational continuity. Laboratory design should include careful consideration of disaster or emergency situations, Detailed planninP g should include such things as emergencyr warning alarm systems, exhaust stack monitoring system, interlaboratory comunications, monitoring equipment and protective clothing located at a site remote from high-risk area, @emergency fire-fighting equipment, evacuation procedures, consequences of power failures, and medical-treatment facilities.

Chapter 2 References 1, Morgan, G. W,, 'Basic Safety Requirements in Radioisotope Work," Paper 20 of a Conference on the Use of Isotopes in Plant and Animal Research. USAEC, TID-5098, 1953 2. Editorial "Cold Setting Lead Cement in Mobile Radiation Shield"t Lead, 20, No. 1, 4, 1950 3. Goertz, R. C. "Fundamentals of General-Purpose Remote Manipulators" Nucleonics, 10, No. 11, 36, 1952 4. Davis, Harold S. "How to Choose and Place Mixes for High-Density Concrete Reactor Shields" Nucleonics, 13, No. 6, 60, 1955 5. Meinke, W. W., Emmons, A. H. and Gomberg, H. J,O "A Versatile Hot Lab for University Research" Nucleonics, 13, No. 11, 1955 6. Disnuke, S, E,, Feldman, M. J o., Parker, G. W,, Ring, F., Jr. "Hot Laboratory Facilities and Techniques for Handling Radioactive Material" International Conference con the Peaceful Uses of Atomic Energy. A/Conf. 8/P/723, 1955 7, Atomic Energy Commission, United States "Volume Three, Reactor Handbook: Engineering' "Volume Six, Chemical Processing and Equipment" U, S. Government Printing Office, Washington 25, D. C. 8. Goertz, R. Co, "Mechanical Master-Slave Manipulator" Nucleonics, 12, No. 11, 1954 9. Ward, D. R. "Design of Laboratories for Safe Use of Radioisotopes" AECU-2226, November 1952 10. Meyer, A. W. "Installing a Small Hot Atomic Laboratory" Industrial Laboratories, p. 62 May 1956 11. Building Research Advisory Board "Laboratory Design for Handling Radioactive Materials" Conference Report No. 3 1951 12, "Control of Radiation Hazards in the Atomic Energy Program," 2.37

2.38 UoS. Atoice Energy Commisston, Wash~, D.C., 1950 13. Skow, R, et al,, "Hazard Evau ation and Control after a Spill of 40 mg. of Radium,,` Nucleonicsq 11, No. 8, 45, 1953 14. Lane, W. et al,, "Contamination and Decontamination of Laboratory Bench Top Materials,,g Nucleonics, 11 No, 8, 49, 1953 15* Ross, DoH,, 0Cleaning Contaminated Surfacese" Soap Sanit, Chemicals, 27, 1951 16. Lane, J. A.o "gContainatlon and Decontamination of Laboratory Bench Top Materials", Nucleonics, 11, NoI 8, 49, 1953 17. Skow, "Hazard Evaluation and Control afteT a Spill of 40 mo. of Radium." Nucleonics, II,, NO. 8, 45, 1953 18. Barry, eto alo, "Radioactive Contamination Sampling by Smears and Adhesive Disks," N cleonlics,;L,. NQo 10, 60s 1953 19. Anoan, "Decontamination Chart," Nucleonics, 9, No. 11, C-12, 1951 20. Breslin, A, J, and Solon, L, R., "Fallout Countermasures for AEC Faciliti$es," NY04682aA,4 Dec, 1955 21. " Radiation and Monitoring Fundamentals for the Fire Service," Published by International Association of Fire Chiefs, Hotel Martinique, Broadwa43l0y at 32nd, Street, N.- Y.. No Y. 1955 22 Curtis, R, L., "Decontamination A Literature Search," Yo964, May 19, 1953, Bibliograph lof ater:ial on decontamination; 70 titles are included, May 19, 1953 23. Rodger, Walton A -and Fineman, Phillip V"A Complete Waste Disposal 'System for a Radiochemical Laboratory"' Nucleonics, 9, No, 6, 50, -1:951: 24. Brown, Ro Eo et al,, "Disposal of Liquid Wastes to the Ground", A/Conf,/p 565, United Nations, NoYo, 9 120, 1956 25. Mawson, C. A,, ""Waste Disposal Into the Grounds," AECL-211 A/Conf.o p12; Atomic Energy of Canada Ltd., Chalk River, Ont, CanaTda United Nations, N.oYo 9 676, 1956 26, Ginell, W, 5, et al,, "Ultimtae Disposal of Radioactive Wastes," Nucleonics, 12, No, 12, 14, 1954 27. Jensen, J, H., "Radioactive Waste Disposa in the Ocean," Nat~ Bur ~ Standards, Handoaok 58, UoS, Govt. Pernt, Office, Washlo DC oC 28, Seli ian, Ho, "The Discharge of Radioactive Waste Products in the Irish Sea, A/Conf/p 418, United Nations, No Yo, 9, 701, 1956

2.39 290 Renna CEo," Disposal of Radioactive Wastes at Sea," A/Conf./p 569; United Nations, No Y., 9, 718, 1956 30. Angel, C. Wo and Ring, F,, Jr., "Wall Transfer Unit and Transfer Carrier for Hot Cells." Nucleonics,, 11, No 9, 69, 1953 31, Argonne National Laboratory "The Argonne High Level Gamma Irradiation Facility,"g Argonne National Laboratory, Lemont, Illinois 32. Atkins, 4, C, and Lorentz, W. N,, "Space-Saving Hot Cella," Nucleonics~, 13, No, 10, 79, 1953 330 Atomic Energy Commission, United States "Fourth Annual Symposium on Hot Laboratories and Equipment," held in Washington, D.C., September 29, and 30, 1955. TID-5280, Technical Information Service, Wash., D.C., 1955; TID-5280 (Suppl 1),9 Jan. 1956 340 Bagnall,, K. W, and Spragg, W" T,, "The Handling of Radioactive Materials - I"B,Atomics and Atomic Technology, 6, No, 3, 71, 1955 35. Braestrup, C, B, and Quimby, E.o H,, "Design and Recommendations for the Radioisotope Laboratory" in Chapter 20, "IPPlanning Guide for Radiologic InstallationsV Scott Year Book Publishers, Inc,, 220 East Illinois Street, Chicago, Illo., 1953 36. Bralove, A. L., "Radioactive Dust Separation Equipment, I, II, and III,," Nucleonics, 8, No, 4, 37, No. 5, 60, No, 6. 15, 1951 37. Editorial, "Adapting Glove Boxes to Gamma Work," Nucleonics,, 7, No, 6, 83, 1950 380 Editorial, "Conoco Gets 'Hot'," Chemo Engng. News, 4753 Octol, 1956 39. Editorial, "Equipment Guide for Radioactivity Laboratories," Nucleonics, 7, No1ll, 90, 1952 40 Farmakes, Jo R,, "Snare-Typ'e Remote Handling Device," Nucleonics, 109 No11, 90, 1952 41, Ferguson, K. Ro, "Design and Construction of Shielding Windows," Nucleonics, 10B No, 11, 46, 1952 42, Fields, P. R. and Youngquist, C, H,, "Hot Laboratory Facilities for a Wide Variety of Radiochemical Problems " A/Conf./p 725; United Nations, No Yo, 79 44, 1956 43, Garden, N. B,, "Semihot Laboratories" Industr, Engng. Chem., (Industr.o) 41, No, 2, 237, 1949 44. Garden, N, B,, "Laborattor3y Handling of Radioactive Material~" A/Conf,/p 722; Unidted Nations, No Yo, 7, 62, 1956 45. Glen, H, o,, "An Engineering Approach to Hot Cell Desi.gn,"

2.40 Proco Amero Soco Civo Eng., 80, No, 446, 1954 46, Goertz, Ro C,, "Fundamentals of General-Purpose Remote Manipulators," Nucleonics, 10, Noll, 36, 1952 47, Gore, To W,, "The New Radiometallurgy Laboratory at the Hanford Atomic Operation," Metalo Progr., 65, Noo 6, 81, 1954 48. Grune, W, N. and Klevin, P, B., "Redesign of a Sanitary Engineering Laboratory to Permit the Use of Radioisotopes,'" Nucleonics, 9, No. 2, 59, 1951 49* Hawkins, M B,, s"The Design of Laboratories for the Safe Handling of Radioisotopes," Isotopes Division Circular B-5 U.S. Atomic Energy Coamission Isotopes Division Oak Ridge, Tenn., 1949 50, Kohl, J. and Newacheck, Ro L,, "Mobile Radiochemical Laboratory," Nucleonics, 10, No. 5, 44, 1952 51. Lane, J. A., "How to Design Reactor Shields for Lowest Cost,".Nucleonics, 13, No 6, 56, 1955 52. Levy, H. A., "Remodeling a Laboratory for Radiochemical Instruction or Research," Industr, Engngo Chemo, (Industro) 41j No, 2, 248, 1949 53. McIntosh, W. W., "Ventilation for Radioactive Work, Balancing, Operation, Maintenance'," Heat, Pip. Air Condit2, 25, No. 7, 98, 1953 54. Manov, G, G,, "Radioisotope Laboratories for Animal and Agricultural Research," in: "The Role of Atomic Energy in Agricultural Research" TID-5115, USo Atomic Energy Commission, Wash,5, D.C, 81 Jan. 1953 55, Man or, G. G. and Biztell, 0. M,, "Design of Radioisotope Laboratories for Low and Intermediate Leveals of Activity," in-: "Syposium on Radioactivt -- An Introduction,"' ASTM Special Technical Publication No, 159 56. Miller, H. S., Fahnoe, F. and Peterson,, W.oR, "Survey of Radioactive Waste Disposal Practice," Nucleonics, 12, No. 1, 68, 1954 57, Monk, Ge S, "Coloration of Optical Glass by High-Energy Radiation," Nucleonics, 10, No, 11, 52, 1952 58. Morris, o,, "An Approach to Hot Laboratory Design," Proc. Amero Soc. Civo Eng&rs.,e 80, No o 448,j 1954 59.e Norris, Wo P., "Radiobiochemical Laboratories," Industr. Engng. Chem., (Industro) 41, No. 2, 231, 1949 60, Obrycki, Ro F,, Ball, R, M. and Da1avidson, W, Co, "Ecnomical Shieldin for Multicurie Sources," Nucleonics, 11, eNo, 7, 52, L953 61, Osso, 0. L. and Gifford, J F,, "Facilities for Dcontaamination of

2.41 Laboratory Equipment," HW26502, Hanford Atomic Products Operation, UoSo Atomic Energy Comtwssion, Wash,, DoC.,, 1953 62, Pravdjuk, N, Vo., "Metal-Research Hot Laboratory," A/Conf./p 673; United Nations, N. YS, 7, 49, 1956 63. Preuss, L. E. and Watsons Jo HL, "Design and Construction of a Small Radioactivity Laboratory," Nucleonics, 6, No. 5, 11, 1950 64. Quimby, E. Ho and Braestrup, C, B,, "Planning the Radioisotope Program in the Hospital,," Amero J, Roentgenol., 63, No.1, 1950 65. Remote Control Engineering Division "A Manual of Remote Viewing," ANL-4903, Argonne National Laboratory, Chicago, Illinois, 1953 66. Rice, Co N., "Laboratory for Preparation and Use of Radioactive Organic Compounds," Industr. Engngo Chem., (Industr.) 41, No. 2, 244, 1949 67. Ring, F., Jr., "Shielding Structure Facilities for Atomic Energy Research," Proco Amer. SOCo Civ, Engrs., 80, No. 447, 1954 68 Ryberg, Jo,, "Shielded Box for Chemical Work," Nucleonics, 13, No, 10, 65, 1955 69. Rylander, E. W. and Blomgren, Ro Ao, "Operating Procedures of a Hot Laboratory for Solid State Tests," Nucleonics, 12, No. 11, 98, 1954 70, Selected Authors, "Hot Labs -- A Special Report," Nucleonics, 12, No, 11, 35, 1954 710 Somerville, A,, "General Motors Builds Radioisotope Laboratory," Nucleonics, 13, No. 10, 68, 1955 72. Spence, R., "An Atomic Energy Radiochemical Laboratory Design and Operating Experience,," A/Confo/p 438; United Nations,, NoY., 1956 73. Steele, R. V., "Remote Radioactive Materials Testing Laboratory at Livermore Research Laboratory," LRL-150, US. Atomic Energy Commission, Wash., D.Co, June, 1954 74. Swallow, A. J., "The Hot Laboratory at the University of Birmingham,," Atomic Scientists News, 1, No, 4, 130, 1952 75. Swartout, J, A,, "Research with Low Levels of Radioactivity," Industr, Engngo Chemto, (Industro) 41, No, 2 233, 1949 76. Tompkins, P. C,, "A Radioisotope Building,t Industr. Engngo Chem,, (Industro) 41, No, 2, 239, 1949 77, Tompkins, P. C. and Levi, H., "ImIpact of Radioactivity on Chemical Laboratory Techniques and Design," Industro Engng. Chemo, (Industro) 4!, No, 2, 228, 1949

2.42 78. Ulm, R.o Wo eat alls "Po8tsvuth Tachnica1 Servicc Buildingut K41L48, U*S, Atom icEnergy Comisson, Was., D.Co. Nov. 29, 1954 79o Waday, Wo Go, "SipleRadiation Shii lding Doos uaoic$, 12, Noo 5, 54, 1954 80O Ward, D9 R, "Deign of Laboratorisa for Safe Use of Radioisotopes," Chaptr~ 8 Rdii s s a i0 RiWenhold Publishing Co., No Yo, 1953 81 Whitalesay, a G Eo. and Gi an, E I"Rea tion Protection of Personnel and Radiochemica1 Lab ratories, T eir Design and Ozperaion," AECU-l020, So $ Atomic Energy Cooeission, Wash., D.C., July 1950 82, Yakovvie ' Goo e t aDlo "A Hot Analytical Laboratory." A/Conf./p 672; United Nations, N. Yo, 7, 57, 1956 83. Arnottq Do Go an d Wel11ls-Col,9 J "A Rapid Method for the Extraction of Radioiodide from Urine," Nature, Lond,, 171, 269, 1953 840 Browder, F. N, "Liquid Waste Disposal at Oak Ridge Natil Lab.,0 Industro Engngl. Chem., (Indusatro 43, 150'9, 1951 85. Carfter o, Wo, "Removal of Radioactive Iodine by Laboratory Trickling Filters," Sewage Idustr.o Wastes, 25, 560, 1953 86. Cowan, F, P. and Nehemiass J. Vo., "Sensitivity of the Evaporation Method of Liquid-Waste Miontoring," Nucleonics, 7, No. 5, 39, 1950 87. Eden, G. Eo, Elkins, G0,oSo and Truesdale,, G.A.,, 'Reval of Radioactive Sub$stances from Water by Biochemical Treatment Processes." Atomics,, 5, 133, 1954 88. Eliassen, R., aufma W.J., Nesbitt, J. B. and Goldman, Mo I,, "Studies ow Radioisotope ReTmSal by Water Traatme t Processes", J, Amer. Wato Wks, Ass., 43, 15, 1951 89. Foster, R. and Rostenbach, Ro,, "Distribution of Radioisotopes in Columbia River,"~ J. Amero Wat, Wks, Ass.,g 476, 633, 1954 900 Ginell, W.4, Martin$ J. J. and Hatch L. P.,9 "Ultite Dispos$al of Radioactive Wastes," Nucleo3 icsr, 12, No. 12, 14, 1954 91. Goldin, A, S., Na der, J $So and Setter, L, R, "The Detectability of Low-level Radioactivity in Water," J. ro Wet, Wks* Ass, s, 45 No, 1 73, 1953 92. Gorman,, A, Eo, "Mutual Interests of the Water Works and Atomic Energy Indus -tries 0" J Amero W at Wks, Asso 43, 86a5, 1951 93. Grue, WO N, a$d Eliassea RB,, "Studies on the Effect of Radioactive Phosphorus on the Biochemical Oxidation of Sewage," Sewage Industr. Waste$ 23, 141, 1951

2.43 94. Hayner, J, Hio, '"Atomic Energy Industry," InduStro Engng. Chem,, 44, 472, 1952 950 Herrinagton, A. C et atO, "Economic Evaluation of Permanent Disposal of Radioactive Wastes," Nucleonics, 11, No. 9, 34, 1953 96. Higgins, E., "Atomic Radiation Hazards for Fish," J, Wildlife Management, 15, 1, 1951 97, Kochtitzky, Oe W. and Placak,, 0. Ro, "How to Survey a Stream for Radioactive Substances,"l Publ, Wks., 83, 76, 114, 1952 98. Loosemore, W. R*, "Monitoring of Water for Fission-Product Contamination,," Nucleonics, 11, No, 10, 1953 99. Monowitz, B. and Hatch, L. P., "Processes for High-Level Waste Disposal," Chemo Engng, Progro, Sumpo Ser, No. 12i, Nuclear Engng,, Part 2, Amero Inst. Chem, Engrs., N. Yo, 144, 1954 100. McCullough, G. E,, "Concentration of Radioactive Liquid Wastes by Evaporation," Industr, Engngo Chem,, (Industrr) 43, 1505, 1951 -101* McKay, H. A. C, and Walton, G. N,, "Safety Criteria in Radioactive Water Monitoring," Nucleonics, 5, 12, Aug. 1949 102. Newell, J, F. and Christenson, Co W,, "Radioactive Waste Disposal," Sewage Industro Wastes, 23, 861, 1951 103o Newell, J, F. and Christenson, C, W,, "What Treatment for Radioactive Wastes," Engng, News-Record, 147, 37, 1951 104. Powell, C, Co and Andrews, H. L, "Radioactive Waste Disposal,t Publo Hlth, Repo Wa sh,, 67, No. 12, 1214, Dec. 1952 105. Rodger, W, A,, Fineman,4 P., "A Complete Waste-Disposal System for a Radio-Chemical Laboratory," Nucleonics 9, 51, 1951 106, Rodgers, W. A. and Fineman, P., "Radioactive Waste Disposal," Chem. Engng., 58, 146, 1951 107. Ruchhoft, Co Co and Feitelberg, S., "Estimates on the Concentration of Radioiodine in Sewage and Sludge from Hospital Wastes," Nucleonics, 9, No, 6, 29, 1951 108. Ruchhoft, C. C,, Gorman, A, Eo and Christenson, C. Wo, "Wastes Containing Radioactive Isotopes," Industr, Engng. Chemo (Industro) 381, 545, 1952 109. Ruchhoft, Co C. and11 Setter, L. Rof, "Application of B iological Methods in the Treatment of Radioactive Wastes," Sewage IndustrO Wastes, 25, 48, 1953

2.44 110o Rudolfs, W,, Ed "dustril Wastes: Their Dis osal and Treatment" Reinhold Coo, N. Y. 1953 111. Setter, Lo R., Goldin, A. S. and Nader, J. S., (Robert A. Taft Sanitary Engineering Center, Cincinnati, Ohio) "Radioactivity Assay of Water and Industrial Wastes with Internal Proportional Counter," Analyt. Chem., 26, 1304,, 1954 112. StraibCC, PG, "Observations on the Removal of Radioactive Materials from the Waste Solutions," Sewage Industro Wastes 23, 188, 1951 113o Strauba C. P., "Effect of Radioactive Materials on Environmental Health," Publ, H1th, Rep., Wash,, 67, No. 3, 1952 114, Straub, C. Po, Morton, R, J. and Placak, 0 R,, "Studies on the Removal of Radioactive Contaminants from Water," J. Aer, Wato Wks, Ass., 439, 773, 951" 115. Western, F,, '"Health Safety Considerations in the Disposal of Radioactive Wastes," Ind. Hyg. Quarterly, 14, No. 3, Sept0 1953 116, Liebe"n, J A,,, "Engineering Aspects of the Disposal of Radioactive Wastes from the Peacetime Applications of Nuclear Technology," J. Am. Public Health Assoc,, March 1957 117. Silverman,.,, "Air and Gas Cleaning For Nuclear Energy Processes,," Proceedings of the International Conference on the Peaceful Uses of Atomic Energy, Geneva, p. 571, 1955; United Nations, New York: 9_, 1956 118. Craig, H., "Disposal of Radioactive Wastes in the Ocean: The Fission Product Spectrum in the Sea as a Function of Time and Mixing Characteristics,",' Scripps Institute of Oceanography, 1958 119. "The Disposal of Radioactive Waste on Land," Report of the Committee on Waste Disposal of the Division of Earth Sciences, Natl, Acad, Sci.-Natl, Res~ Council, 1957 120. Rhodes, D. W., "Waste Characteristics Goveming Fixation in Soils," USAECReport TID-7550, Hanford Atomic Products Operation, 1958 121. Gri tt, E S., "Calcination of Aquaenous Reactor Fuel Wastes in a Fluidized Bed," USAEC Report TID=7550, Phillips Petroleum Co., 1958 122. Thomas, R, G. and Christenson, C. W "Leaching Studies on Fired Clays Containing Radionuclid$es," USAEC Report TID-7550, Los Alawms Scientific Laboratory, 1958 123. Goldman, M. I,, Servizi, J, Ao., Lauderdale, R. A.o and Eliassen, R., "Fixation of Fission Protgucts in Ceramic Glazes,' USAEC Report TID-7505 Massachusett$ Ins$titte of Technoloy, 1958 124. Gloyna, E. F., Schecter, Ro and Serato, So, "$Storage of Reactor Fuael Wastes in Salt Formations," USAEC Report TID-7550, University of Texas, 1958

CHAPTER 3 FILM, GLASS, CHEMICAL AND CALORIMETRIC DOSIMETRY The term "dosimetry" originated in the medical profession and involves the methods of the determination of the "dose" of radiation absorbed by a body or material placed in a radiation field for a given period of time. In general, dosimetry is concerned with the measurement of radiation dosage by the chemical or physical change the radiation produces in a measuring system. Examples of such a system are a photographic plate which is "exposed" by the radiation, or a chemical solution in which a cation is oxidized or reduced in the presence of radiation. The early use of X-rays and radium in cancer therapy introduced the need to measure the radiation dose given to patients. Ele.ctronic instruments of advanced design such as described in ChapterllJ*..wwe not available at that time although simple forms of the ion chamber were used extensively. Numerous chemical and biological systems were explored for use in dosimetry. Crude photographic films were available and were used in the early 1900's to measure radiation dosage. For convenience in treatment the dosimeters will be discussed under the following categories: A. Film and Plastic Dosimeters B. Glass Dosimeters C. Chemical Dosimeters D. Calorimetric Dosimeters E. Miscellaneous Types of Dosimeters The first three of these systems depend on chemical changes to detect radiation whereas the fourth measures the heat effect of the absorbed radiation in a substance of known heat capacity, and hence depends on a physical change. 3.1 Film Dosimeters In addition to the silver halide photographic-type film, plastic films containing dyes have been used in various dosimetry studies, but much less extensively than the silver halide films. 3.2 Film Badges Silver halide films are used in every radiation laboratory in the film packets of personnel film badges. In the routine operation of any radiation laboratory or area of significant radiation level, records must be kept of the radiation dosages received by all personnel having access to *Radiation Uses in Industry and Science.(B7) 3-1

3-2 the radiation areas and this is usually accomplished by the use of film badges. This is essential to prevent excessive and harmful exposure of the personnel to radiation as a result of continual working in an area of high radiation level. Also, such records provide protection from legal action by either employees or visitors who unjustly claim over-exposure to radiation. Tolerance levels and the general problem of health physics are discussed in Chapter 1. Figure 3,1 shows an employee in a radiation laboratory inserting a film packet in a badge. Note that he is also wearing a film badge on the lapel of' his laboratory coat. The satme badge is worn daily by laboratory employees and the film packet is changed periodically, usually once or twice per month or immediately after a radiation incident of suspected over-dosage. The exposed film packets are returned to the supplier where they are developed and returned with a report of the accumulative dosage received by each badge. Nuclear radiations and X-radiation darken the X-ray type films that are used in the-badges in a manner similar to the darkening of photographic film as a result of the exposure of the film to light. Both the X-ray and ordinary photographic fil~ms consist of a light-sensitive emulsion layer deposited on a cellulose acetate or glass base. The light-sensitive emulsion layer consists of a colloidal dispersion of crystals of a silver halide (or a mixture of silver halides) in gelatin. According to the Gurney and Mott theory,(l) light or ionizing radiation incident upon a halide crystal will transfer some energy to the crystal, causing one or more of its electrons to reach conduction energy levels. These mob'ile electrons then move about in the lattice until they are trapped by Centers of a deformity or impurity. The coulomb forces set up by the trapped electrons cause silver ions in the crystal to migrate to those electrons where the ions are neutralized to form silver atoms. The silver atoms form the socalled photographic "latent image" and catalyze the reduction of the entire crystal to metallic silver when the film is developed. If overdevelopment is avoided only those crystals that have been activated will be reduced to metallic silver. The unreduced halide is then removed by forming a watersoluble complex with thiosulfate ion in the fixing bath. After fixing, the film is washed and dried and is ready for use. Figure 3.2 shows the range and sensitivity of some typical emulsions to gamma radiation from a radium source.(2) All emulsions show a great variation in sensitivity to the energy of the activating X- or gamma-radiation. In practice this is taken into

Figure 3.1. Insertion of Film in Film Badge.

0.2 3 6 Eastman Type K Kodak Liquid 0. 2 3 Du Pont 502 X-ray Developer 0.2 3 6 Ansco Non Screen Developed for 5 mm. at 680F 02 3 g Ansco Super Ray 0.25 3 5 Eastman Type A 0.3 3 6 Du Pont 510 0.5 3 3.7 Eastman Cine Pos. 5301 DuPont 605 0.3 3 4.8 Eastman Cine Pos. 5302 0.3 3 4.1 Du Pont Defender Adlux 0.25 3 Ansco Reprolith Ortho 0.5 3 4.9 Eastman 548-0 Double Coat 0.2 3 4.5 Eastman 548-0 Single Coat 0;2. 34.7 I I ~~~~~I P.1 o5 0.1 1 10 102 l03 l04 l05 Exposure (r) Figure 5.2. Range and Sensitivity of Typical Film Emutlsions Used in Film Badges. (2)

3-5 account by covering the film with filters that reduce the sensitivity of the film and flatten its response to radiation of various energies. This effect is illustrated in Figure 3.3 for which the "sensitivity" is defined Z2) as "the ratio of radium gamma ray exposure to the exposure at a given energy, in roentgens, which is necessary to produce the same film dens ity." For a given film the degree of darkening is-dependent upon the properties of the film and the type, duration and intensity of the radiation that retches the film. To cover a wide radiation dosage range, two or more films are usually used as indicated for one type of film packet shown in Figure 3.4. If two films are used in the packet one of the films is usually highly sensitive and may cover the range of radiation dosages from 0 to 2 roentgens. The other film usually is less sensitive and may cover the dosage range up to 30 roentgens (100 times the maximum tolerance for one week). The two films are wrapped together in thin opaque paper to keep out light which would expose the films. In one type of badge shown at the right in Figure 3.3, a portion of the film is located at a "window" and a portion behind a thin cadmium shield. In the case of exposure to mixed radiation (beta and gamma) the beta particles are stopped by the cadmium shield but the penetrating gamma photons pass on through and expose the film. Beta particles readily penetrate the film packet at the window and expose the film but do not penetrate the cadmium shield. In another type of badge shown at the left in Figure 3.4 a stepped copper wedge is used in place of the cadmium. This type of badge is particularly useful in determining dosages resulting from exposure to X-radiation. The film density varies greatly with the kilovoltage used in the X-ray machine. The use of copper steps permits the comparison of the film density under each step with standards exposed to known dosages of X-radiation produced by machines operated at known kilovoltages. If there is a possibility of neutron exposure, a third film which is sensitive to proton radiation is placed behind the "beta-gamma" films as indicated in Figure 3.5. With a film badge such as shown in Figure 3.5 fast neutrons pass through both the window and the cadmium barrier but slow neutrons are readily captured by the cadmium and, therefore, pass only through the window. The fast neutrons interact with hydrogen atoms in the film to produce recoil

I- ~~~~~ ~3-6 >- 30 >z 20 LAt Unfiltered z 2O U~3 8 O ).0 20 "B rass 0 040Cadm0.020"Cadmium (I) i- D, 0|~.-040"Cadmium LLJ L 0.8 0.6 0 O.4 0 50 100 150 200 250 300 350 400 EFFECTIVE ENERGY, KEV Figure 3.3. Relative Sensitivity vs. Effective Energy for duPont 502 Emulsion and Filtered X-Radiation. (2)

COPPER STEP WEDGE IDENTWICAtION CADMIUM FILTCR *rcaIIOM SHIELD IWOri LEAD SACKING 2 FILMS 2 FILMS Cross Section View of Cross Section View. Of Xoray Badge Radioactivity Badge Figuere 3.4. Cross-Section of X~-Ray and Beta-Gamma Film Badges (Courtesy of Tracerlab, Inc.).

BETA GAMMA PACKET (2 FILMS) i\ "\ "\ ' \" I I I P ' \ I, CADMIUM FILTER NEUTRON PACKET (1 FILM) LEAD IDENTIFICATION SHIELD Figure 3.5.Cross Section of Neutron Film Badge (Courtesy of Tracerlab, Inc).

3-9 protons that expose the film. The slow neutrons also produce protons from the N14 (n,p) C14 reaction, giving an additional proton exposure.. The protons produced by both processes leave "tracks" on the third film. The dosage for fast and slow neutrons can be estimated by counting the tracks after development of the film. This involves microscopic analysis and use of the phase microscopy principle. One commercial laboratory(4) uses Eastman Type NTA film with an emulsion thickness of 25 to 30 microns, calibrates the film with polonium-beryllium neutron sources, and analyses the film at 860X. The beta-gamma dosages are obtained by comparison of the film density at the window and behind the cadmium with that of calibrated control films exposed to known dosages of radiation. To avoid errors from variables in developing, control films are developed and fixed with the test films. Table 3.1 summarizes commercial film-badge services that were available in 1955. The table does not include the AEC operated film-badge service available only to contractors of AEC, nor a listing of neutron-badge services. Additional information on dosimetry with silver halide films is given in References 6-32. 3.3 Polyvinyl Chloride Films The application of polyvinyl chloride films to gamma-radiation dosimetry measurements was suggested b Steigman(33) and investigated in greater detail by Henley and Miller.(5 ) The reaction is thought to take place by the action of gamma radiation on the plastic to liberate HC1 within the film. If an acid-alkali indicator has been incorporated into the film, a color change will occur under the influence of the HC1. Henley and Miller give the following four requirements that a system must meet for acceptance as a dosimeter method: 1. The system must vary in an exact manner and be independent of radiation intensity, wavelength, temperature, pH, etc. 2. The reaction must be irreversible. 3. The results must be reproducible within a few percent. 4. The chemical system must be stable under ordinary storage conditions. Polyvinyl chloride films were studied also by Welshans in the Fission Products Laboratory, The University of Michigan, to check their use as a gamma-radiation dosimeter.(35) Briefly their procedure involved the

3-10 TABLE 3.1 SUMMARY OF COMMERCIAL FI-M-BADGE SERVICE (Courtesy of Nucleonics, Feb., 1955)(5) Minimum Cost of How Often Control Time To Interval Firm Name and Service No. of Subscript. Are Badges Badge Report Badge is Address (per badge) Badges Period Supplied Provided Exposures Kept on File Atomic Research Lab., $1-1, 80$ -2, None None Weekly, bi- Yes 7 days Permanently 2633 Santa Monica 650 -3-10, Blvd., Santa Monica, 60~ - 11-25 Calif. Isotopes Specialties 50~ apiece 13 None Weekly Yes Within 6 years Co., 3816 San Fer- (quantity 7 days nando Rd., Glendale, discounts Calif. allowed) R. S. Landauer, Jr. $1.20-1, 50~ None 13 weeks Weekly, bi- Yes About Permanently and Co., each add. (shorter weekly, or 2 days* P.O. Box 102, Over 50- at higher special arPark Forest, Ill. quantity price) prices Nuclear Consultants, $1-1, 60o - 1 3 months Twice a No 7 days 2 years Inc., 2-9, 50~ - 10 month 33-61 Crescent St., and over Long Island City 6, N. Y. Nuclear Inst. and 65~ -3-24, None None Weekly, bi- Yes 4 days* Permanently Chemical Corp., 60~ -25-49, weekly, or (3 or 223 W. Erie St., 55~ -50-99, monthly more Chicago 10, Ill. 50~ - >10Ot badges) Radiation Detection 50~ (weekly) None 26 wk. Weekly, bi- Yes Within Permanently Co., 576 College Ave., 60~ (bi-wk.) if 1-2, weekly, or 2 days* Palo Alto, Calif. $10 per yr. 13 wk monthly (monthly) if 3 R-C Scientific 60~ apiece None 13 Weekly or Yes Within Indefinite Inst. Co., 307 $2.50 min. periods longer 7 days Culver Blvd., quantity scheduled Playa del Rey, Cal. discounts period St. John X-Ray Lab., 50~ per None 1 year As desired Yes Within a Badge is reCalifon, New Jersey badge user holds few days turned to user $75/yr min. supply Technical Associates, 60o$; less on 3 13 Weekly, bi- Yes Within Indefinite 140 W. Providencia quant. or yr weekly, or 7 days Ave., Burbank, Calif. contracts especial arrangement Technical Operations, 75i each None None As ordered Yes 3 days Indefinite Inc., 6 Schouler Ct., (discounts (3 mo. max.) (sold) Arlington, Mass. on contracts) Tracerlab, Inc.,* 60~ apiece, 3 13 weeks Weekly or Yes Within Permanently 130 High St., over 25- any regu- 5 days* Boston 110, Mass. quantity lar schediscounts dule * Will notify of any overexposures by collect telegram, if requested. Deduct 5% for yearly contract on weekly basis. i For the 17 western states; Tracerlab, Inc., Western Division Film Badge Service, 75% 23rd St., Richmond 2, Calif.

3-11 addition of powdered resin to a filtered solution of dye in chlorobenzene and heating the mixture for 1 hour at 1100C. The solution is then poured on level glass plates and set aside until dry. Welshans reported(35) that films cast from cyclohexanone were clear and free from surface imperfections found with chlorobenzene as the only solvent, with the additional advantage that stock solutions of uniformquality plastic can be prepared and cast at room temperature. However, the drying time was increased to 9-19 days instead of the 1-3 days for chlorobenzene. Hence, in practice about 40 percent chlorobenzene was added to the solvent to decrease the drying period without affecting the appearance of the film. A typical stock solution has the composition: 200 c.c. chlorobenzene (tech) 300 c.c. cyclohexanone (Eastman) 0.11 g methyl violet 6B (General Aniline) 23.7 g polyvinyl chloride resin (Geon 101) The dye is dissolved in the mixed solvents and resin added slowly to avoid lumps. The mixture is heated to 1150C for 1 hour, cooled, and filtered through cotton into a dark-brown bottle. If 76.5 cc of the solution are poured on a 19 x 29 cm glass plate, the dried film is from 0.0016 to 0.0018 inch thick. These films were cast on an 8 x 12 inch glass plate which had a wall of sauereisen cement around the edges to act as a dam for the solution. The plate was floated in a pool of mercury contained in a large pyrex dish, to keep the surface of the plate perfectly level. The plate containing the dried film is placed under water and a edge of the film lifted with a sharp knife. Then the film can be peeled off in one piece without tearing. It is finally dried with blotting paper and cut into the proper shape with a razor blade. The stability of the film was demonstrated by boiling it with water for several minutes. The appearance and shape of the film remained the same after the test and the water remained colorless, although the dye is normally water-soluble. It is necessary to provide some type of support to hold the films during exposure or reading in the spectrometer. One method is the insertion of the film in a 35-mm Kodak ready-mount.

3-12 In Welshans' studies, the films were placed in a holder and transmission readings taken before and after exposure to gamma radiation with a Beckman Model DU spectrophotometer kept at a wavelength of 600p. and a slit width of 0.80 mm. The tungsten lamp was used in all measurements with the Beckman filter in the "in" position and the phototube in the "out" position. The sensitivity of the instrument was adjusted to give 100 percent transmission in air without the film in the holder. Using this technique it was possible to make readings on the same film that agreed within 1 percent. If the reciprocal of the exposure time is plotted versus the percent transmission, a value of the transmission, Tinf, is obtained when the curve is extrapolated to infinite time as shown in Figure 3.6. Welshans reported-that unfiltered (250-kv) X-radiation is about 2-1/2 times as effective in bleaching as the cobalt-60 radiation for equal dosages. This is in agreement with previous findings and indicates that it is not possible to use the films to compare various sources of gamma radiation and that each wavelength must have its own calibration curve. As a result of these observations Welshans(35) made the following conclusions regarding the use of PVC films as dosimeters for gamma radiation: 1, The system does not vary in an exact manner. It has been shown to be a function of wavelength, pH, and possibly intensity at high levels of radiation. The effect of temperature was not studied. 2. The reaction is irreversible as carried out in dosimetry methods. 3. The results indicate that they are reproducible, over a limited range, within 5 percent. 4. The system is stable if stored out of sunlight or bright artificial light. Sunlight will almost completely bleach the film in two days. A No. 2 photoflood, 2 feet from the film, will cause a 10 percent increase in transmission in 20 minutes. However, it appears that limited exposure to normal levels of artificial light has no effect. If the PVC films are used in a gamma radiation field of constant energy, such as cobalt-60, and the total exposure is kept below 10 million roentgens, they provide a convenient and reproducible method of measuring dosages.

0 I0 20 E30 X _J /' 1L40 0 Z 50 o o (0 o X-I-KILOCURIE COBALT-60 SOURCE (0 o0 0.01 0.02 0.03 Q04 RECIPROCAL OF RELATIVE DOSAGE Figure 5.6. Percentage Transmission of PVC Film vs. Reciprocal of ~Relative Dosage fromIE Cobalt-60 GamSOURas.E e:70 -90 I00,-, 0.01 0.02 0.03 Q04 RECIPROCAL OF RELATIVE DOSAGE. Figure 3.6. Percentage Transmission of P'VC Film vs. Reciprocal of Relative Dosage from Cobalt-60 Gemtuas.

3-14 3.4 Cellophane Films The principal disadvantages of the PVC films were the long exposure time required, the difficulty of preparing uniform, good quality film in the laboratory, and the necessity of calibrating each sheet of film. Also, the hand made film is quite expensive because of the man-hours involved in its preparation. Henley(36'7) recently reported the use of a moisture-proof, heat-sealable cellophane containing a dimethoxy-diphenyl-disazobis-8 amino-lnaphthol-5, 7-disulphonic acid dye.(8) The film is quite inexpensive, being priced at only 0.042 cent per 1000 sq in. The physical and optical characteristics are uite uniform. The commercial film identified as duPont No. 300 MSC(39 has a controlled thickness of 1 mil and an initial transmission that is fairly constant.(36) In using Henley's technique, sheets of cellophane are cut into strips about 1-3/4 inches long and 1/2 inch wide. The sheets are then placed in an aluminum holder especially constructed to fit the standard Beckman cell holder. Henley reports that the most satisfactory absorption peak was found at 6550 Angstroms. This wavelength was used with a DU Model Beckman spectrophotometer with a slit width of 0.15. Henley exposed cellophane films of the type described to varying dosages of gamma radiation from cobalt-60. In all radiation exposures the film was sandwiched between 1/8-inch-thick sheets of polyethylene. The changes observed in percent transmission as a function of radiation dosage are shown in Figure 3.7 and difference in percent transmission as a function of radiation dosage is shown in Figure 3.8. Figure 3.8 shows calibrations at dose rates of about 100,000 roentgens per hour from cobalt-60 sources and includes data from the large gamma source (40-42) in the Fission Products Laboratory, The University of Michigan. According to Henley, the dye in the film is decolorized as a result of reduction by radiation in a statistically random manner. A given fraction of the dye is reduced to the leuco form following a first-order chemical reaction. Color is not regained and there is no subsequent darkening of strips stored over 1 month. Also neither the addition of acid or base nor contact with oxygen will produce a return of the color.(37) An equation for the curves of percent transmission vs. dosage may be written as: r = (0.68 +.01)(106)(T - To) (3.1)

PER CENT TRANSMITTANCE -~- N- -N- PO P r P N N - N N V N 01 - OD f 'I ID ') I '' 0 0 x X G) cOD Figure 3.7. Percent Transmission vs. Gamma-Radiation Dosage for Cellophane Film. (36)

F- -- LI ' '. l, IIlllI 12 H 8 -CD O 2 4 6 8 10 6- c~ QIH4 -~~~~0 0 2 4 6 8 10 DOSE IN ROENTGEN X 10-6 Figure 3.8. Change in Percent Transmission vs. Gamma-Radiation Dosage for Cellophane Film. (37)

3-17 where: r = dosage in roentgens T = percent transmission at dosage r To = percent transmission at zero dosage According to Henley, the probable error of the calculated dosage over the range shown in Figure 3.7 varies from about 7 percent at 3 x 106r up to 60 percent for dosages as low as 200,000 r. However, at higher dosages for which the transmittance change is more than about 20 percent the decomposed dye molecules begin to compete with the dye for radiation and the kinetics of the reaction are changed. In this range Henley recommends plotting the log of absorbance vs. dose. Such a plot gives a straight line indicating that destruction follows the "direct hit" or "target" theory. A comparison of the results obtained from cobalt-60 gamma radiation with those obtained from high-speed electron bombardment using accelerators is shown in Figure 3.9. Note that the change in transmittance for a dose of 106 roentgens is independent of dose rate but that those for electron radiation are 2.2 times greater than for cobalt-60 gamma radiation. Henley suggests that this difference is explained on the basis of the difference in linear energy transfer (LET) between cobalt-60 gamma photons and 2-Mev electrons.(37) Figure 3.9 shows Henley's calibration curve for electron radiation. (37) 3.5 Coloration in Plastics The polyvinylchloride (PVC) film dosimeter described previously incorporates an acid-base indicator. However, the color changes in PVC itself and in other plastics such as polymethyl methacrylate and polystyrene could possibly be used for dosimetry. The coloration of PVC by radiation is proportional to absorbed dose and Artandi(43) has suggested its use as a dosimeter in the range of 105 to 107 rads. In polymethyl methacrylate radiation causes changes in both color and ultraviolet absorption. Fowler and Day(44) have investigated these changes in both polystyrene and polymethyl methacrylate. However, Artandi(45) concludes that according to their findings neither effect in polymethyl methacrylate is usable for dosimetry as nonlinearity was observed in the range 1 - 6 x 10 rads. 3.6 Chemical Changes in Plastics In addition to color changes as described above use of certain chemical changes produced by irradiation of plastics have been investigated for dosimetry. These changes include polymerization, degradation, and gas evolution.

(0 '01 Lii (9 Ui4 z 2 I- x-x -x \Oj K I0 16 1 108 0 DOSE RATE, ROENTGENS /HOUR Figuxe 5.9. Comparison o~f Gammua Photon and Electron Radiation and, Different Dose Rates ~for Celiphane Films (Semi-.Log Scale). (57)

(a) Polymerization The effect of radiation on various polymerization reactions have been studied. Such systems are dependent on radiation intensity and on impurities and the techniques of measurement are rather complicated, and hence are not very practical, for dosimetry. (b) Degradation Alexander has studied the degradation of solid polymethyl methacrylate by ionizing radiation.(46) He suggested use of intrinsic viscosity measurements of the degraded polymer as a method of dosimetry. The reciprocal of viscosity molecular weight varies linearly with the dose over a hundredfold change in molecular weight or over a dose from a fraction of a megarad to 100 megarads. Feng has proposed a dosimetry method based on degradation of polystyrene in carbon tetrachloride solution.(47) (c) Gas Evolution Certain gases, particularly hydrogen, are evolved upon irradiation of polymers. The evolution of hydrogen is associated with free radical or double-bond formation and cross linking. The dehydrochlorination of polyvinylchloride and its use as a method of dosimetry is already described. The amount of gas evolved can be used as a measure of the dose received by a polymer. However, difficulties are encountered because of slow diffusion and dependence of the rate of evolution on sample size and shape. 3.7 Glass Dosimeters Irradiation of glass causes changes in the absorption characteristics for both ultra violet and visible light. These changes have been used as a basis for several types of glass dosimeters. 3.8 Phosphate Glass Personnel Dosimeter A phosphate glass dosimeter based on the principle of radiophotoluminescence was developed as a personnel dosimeter for the U.S. Navy.(48-51) The dosimeter is known as the DT - 60/PD and is used to measure radiation

3-20 exposure levels received by military personnel in the same way that film badges are used for laboratory personnel. The waterproof glass dosimeter is much more rugged than the film badge and is useful over a range of 10 r to 600 r. In the energy range from 5 Mev down to 80 kev the radiation dosage can be measured within an accuracy of + 20 percent (except for the range of 120 to 150 kev). The glass used is a silver-activated phosphate glass of relatively low cost and adapted to large-scale manufacture. The phosphate glass base has(49) a composition by weight of: 50 percent Al(PO3)3, 25 percent Ba(P03)2 and 25 percent KP03. To this base glass, 8 percent AgP03 is added to produce the optimum properties for dosimetry. This silver-activated phosphate glass emits a fluorescence when it is exposed to light in the violet and near-ultraviolet region. The intensity of the fluorescence peaks sharply at about 2400 A as shown in Figure 3.11. When the glass is irradiated the peak at the shortest wavelength is shifted so that the glass no longer fluoresces appreciably at 2400 A but requires a longer wavelength as is indicated in Figure 3.11 by the dashed line peaking at about 3300 A. The increase in intensity of the fluorescence at the longer wavelength is directly proportional to the radiation dosage received as shown in Figure 3.12. It is believed that electrons produced by the interaction of radiation and glass are trapped by Agt ions in the glass, reducing the ions to "atomic" silver centers, AgO, which are responsible for the new absorption band.(49) The fluorescence is also a function of the radiation intensity. The dependence upon the energy of the radiation is reduced by placing two lead shields on either side of the glass with a small hole (0.107 in. diameter) to improve the response to the lower energy radiation.(50) This corresponds more or less to the window in the film badge. The increase in fluorescence at the longer wavelength is permanent. This is an advantage as it permits keeping glass dosimeters on file over a long period of time. Storage for months at room temperature does not affect the sensitivity of the dosimeter. The dosimeter itself is a locket type which is designed to be worn around the neck. A block of the glass 3/4 x 3/4 x 3/16 inch is enclosed in a circular black plastic case 1/2 inch thick and 1-1/2 inches in diameter. The case is made in two threaded mating halves which are firmly screwed together to hold the glass as shown in Figure 3.13.

A\ % TRANSMITTANCE o ~ A ~ ~ o ~ a ~ O m I As I rO -I 0E O0 0~ — IT) m Figlure 7.10. Calibration Curve for Electron Radiation of Cellophane Films. (37) Oo-D m - G) ' ~ m Figuire 5.10. Calibration Curve f~or Electron Radiation of Cellophane Films. (3.7)

Excitation / of I of,, >-80 Ag,-ions \ Emission of / __ Excitation \ A\missiont of (n I // Ag0 centersgo 2P00 200 3,600 4,400 5,0 6,000 6,800 WAVELENGTH (ANGSTROMS) Figure 3.11. Spectra of Nonirradiated (Ag+ Centers) and Irradiated (Agr) Phosphate Glass. (49) Phosphate Glass. (49)

500 400 ULU Lu O300.200 100 0 100 200 300 400 500 APPLIED X-RAY DOSE, ROENTGENS Figure 3.12. DT-60/PD Dosimeter Response. (49)

U) H to to -' ---- to U)rd rdH HU)O ci 11 to q- p1 ocDg C) rd r003 U) to H .C'4 %r'c..H C) toU)03 0 H 0 ON.....,..... to bD.-ito -- Q30-1 0 O.\,1 H U) U)Cf2rQ -Pro U) r4toOto -H to 4 t 0-H 9 V H U) 10

3-25 To read the dosimeter the locket is opened with a small spanner wrench type tool (see E of Figure 3.13) and the glass block is removed and placed in a special instrument called a "fluorimeter" or "dosimeter reader". Here the glass is exposed to 3650 A light and the fluorescence read directly from the calibrated scale of the meter. The longer wavelength is used to isolate the AgO emission band from the Ag+ emission band. A photograph of one model of dosimeter reader is shown in Figure 3.14. 3.9 High-Dosage Phosphate Glass Schulman(52 53) has reported that the phosphate glass used in radiophotoluminescence dosimetry by the U.S. Navy also can be used for measuring much higher dosages than the 600 r limit for personnel dosimeters. For multimillion rep exposures, another property of the same glass has been found useful as a dosimeter. Changes of optical density, under standard time and temperature treatment, have yielded simple, inexpensive, and reproducible means of measurement of radiation exposure. Figure 3.15 shows a photograph of the glass blocks before and after irradiation. After irradiation the glass develops a broad absorption band which peaks at 3200 A and then spreads out into the visible region and which gives a yellow-brown coloration to the irradiated glass. The optical density as a function of wavelength is shown in Figure 3.16 for phosphate glasses with and without silver before and after irradiation. Figure 3.17 shows the dose dependence of the absorption of phosphate glass measured at different wavelengths. At low dosages, adequate optical-density measurements are made at 3500 A using 3 mm thick glass. However, at higher doses and 3500 A the optical density is too great and a longer wavelength should be used. The optical density resulting from a given dose to glass or film, as determined by a primary method, such as ferrous sulfate dosimetry, is found to vary sharply with radiation energy. Thus, although this method provides, after calibration, a useful tool for a particular source and geometry, care must be taken in applying data from one laboratory to the work of another. Figure 3.18 shows the variation in phosphate glass sensitivity to X-radiation of different energies relative to its sensitivity to cobalt-60 gamma radiation. Figure 3.18 shows that the sensitivity between 200 kev and 1 Mev is quite constant. This is the range of energies of gamma photons for most

3-26 Figure 3.14. One Type o f Reader for Phosphate Glass Dosimeter. (.9,50) Figure 3.14. One Type of Reader for Phosphate Glass Dosimeter. (49,50)

:::_:-: i_:::_:-:::::::::::::;:::: i:iii:i:-iii-li-iiici'iiiliili:i:iiiiii ~i:::ii:i:::::i::i':i:i i:i:::::::::i:-:i:1:i:::::::::::::::::-:::-:-::::~: I''i:''':'::8'::''''':i;iii.:j'i:: '''''''':::':::':::' riii:ii'ii:ii-,iiii'ii-i:i:-iiiii:ii)iii:::::::,i.i:iii:::i:i:: '::':': -: ':: -- -:':::: —:-::-::':';::':: i:i~i:i:- iii iiiii:'iiiiiTi,,i,,i,,jiiiiii:i:iii:::::::::::i:i-ii'i i'iiiiiiii"ii:'iii''iii-i:.ciiiii-i:i-i::-:i-i:i~ii-:-:-:~::- -::.:. i: _i:-i:i:ii —i:ii-~i:ii ~-i:::-:i:i:ii:i:::~~'::i:::::::~:i:i::::::::i~i:-ii::i:_:i —:i:::-:::: ':::::::i::-::i:::i.:::::::..,:: ':'ri:i:liii::-:'':::-:-:: i:::i:-:i:ii-: —:- I: —:-: I:i: -.::.::::::_-~......:..::i:::::i::i:::::-:::~:,:-i:::::::::::::::~::ii:::i::.:ii:iii:i:i::::::::~:::: -:iiiiiil-iiiiiiii iii-iiii~ii:ii:::::::::::::i;:i;i'~l;: ~';::ii:i'ii~iiiiiiiiii~iiii~:i iiiiiiiiiiiiiii:li-iii.i:i:i.i:i:::::-::-:-_:-::::::::::::::::::~ -iiiii-i:i:i:i:i:i:ilaisiiiiii::ii';:i~:i::.:il-:i:. i'iiliiiii~iiili':i:rli-iiiiii:i~i-'iii ':: -i: ':-'-:-::::::::::-:::::::. -:-: -:-:::,:a:;:-::::::-:i_~::::::::i-i:iiiiii:i-i ~-:;::::-::::.:-:::::::::::::~::.;:.... i:-::::::: i:i:i:i:::i:i-iiliii~i:i:-i:i:::::::::::: —iii::i''-i'i:i'i~i-i'i -i- ~: '::::::::.::i:ii~-:i:iiiii:i:::-:i-:i:~:::::-: -:::i:i:I:i:iiii:::i~:::::~:i:i:i:i:i:i::::::::~:::::::::-:-:::::::: i-i —:i:i:i:i:iii i:ii:... i iii' ~iiii i-: —:i-i:i:i i.:..: i'ii: iii-iii:'::.:::":::':-:: eailr i:::.:-: ~::::: 1:::::-:-:::_:::_:-::: iiiiXliiii iiiii~ii:i:i'ii —,i:-:_:i:i:::-::::::::::::::::::::::~:i:i:::i:.:iiii:iii iii:iiiiii:i:i:i~:~::::~:::::::::::,::::::::::::::-:i.i:::i:i:_:::::-::::::.:...:::::: ii'.ii:iii-i-i:iiiiiii:iii iiiiiiiii:ii iii-ii -:::i:ii:i -::;:_:i-::.: ~;-~~~~:~:::i:-;i:~l:i i:::::::: i::::i:_:::::::::.::::-::::i:::i:ii-:-.i:iiiil:i::.:::-:-::::::-:-::: i'i'iifi'~ii~i'ii~i:iii~i:i.i — ii:iiiiii:ii iiiii ii:iiiiii;:i -.:::::::.::::::::::: -::.:::::.:::.-:.-::::::::: ---::.:::-::::::::::::::.:::::-::::.::::::: -,-;:: —: —:::i:i:i:.:i:::::::i:::::.::i:i:l:::i::::~~:i:-::::::::::: iiiiiiiiiiii -iiiiii:i'i~:':::::::::::: iji:iiiiiiiiiiiiii il-:i::::::::-:-:::::~:~:~:::::::~:~;:i:::::::i:i:i:::i:::::: —: -::- ---:::::::::::::::::::::i:i i:i:::::::::::::::::;:~~:~::::::i:::::::- i::::::::: iiiiiiiii:-iii:iiiii:iil -:::::::::::::::: i~~L~:i'~.:~':rl;:':;~l~~i iiiiiiiiii-i:iLhi I tU::::::::':':':: Figure ,1. Photograph of Phosphate Glass Blocs Before and After Irradiation.

3.0 PHOSPHATE GLASS 2:.5 ~~2.5 tI =WITH OR WITHOUT Ag, BEFORE X-RAY F 2.O0 X 2 NO Ag, AFTER X-RAY en | \3 - WITH Ag, w 0 1.5 mI 0 ' 3 ~I.O i.2 0.5 0 3000 4000 5000 6000 WAVELENGTH (ANGSTROMS) Figure 3.16. Absorption Spectra of Phosphate Glasses. (53)

10.0 I.z w IL. U. U. 1.0 4 -_.,4500A *~ 8 BROOKHAVEN EXPOSURE + MIT EXPOSURE.10 * BROOKHAVEN EXPOSURE * MIT EXPOSURE BROOKHAVEN EXPOSURE ~ MIT EXPOSURE.01 1/3 10, lo..lo.. lop EXPOSURE (ROENTGENS) Figure 5.17. Dose Dependence of Absorption of Phosphate Glass Measured at Different Wavelengths. (55) (log-log scale)

=I. 18 LLJ LL LU 0 50 100 150 200 250 300 1200 EFFECTIVE ENERGY (KEV) OF HEAVILY FILTERED X-RAYS (Go60) Figure 3.18. Dependence of Phosphate Glass Sensitivity Upon Energy of Radiation. ( 53) mL -JI t.~ q ~ ) o of Radiation. (53)

3-31 radioisotopes of commercial interest; such glass dosimeters should therefore be useful for radiation spectra in this range. Another problem is the tendency of the glass to fade with time. This effect may be reduced by heating the glass to 1300C for 13 minutes after exposure. This accelerates the removal of early fading without destroying the more stable color centers. A comparison of the fading in heat treated and non-heat treated glass is shown in Figure 3.19. Using optimum procedures, dosages up to 2 x 106 rep can be determined within an accuracy of 5 percent. At higher dosages the curve of density vs. dose becomes so flat (see Figure 3.16) that the accuracy is decreased. A special development is the small volume dosimeter shown in Figure 3.20 which may be inserted in the body to measure doses used in radiation therapy.(55) 3.10 High Dosage Cobalt Glass A "cobalt" glass has recently been developed(56,57) that shows considerable promise for use in high level dosimetry. The glass has the approximate composition of: 62% SiO2 11% Na20 21% B203 6% A1203 0.1% Co304 This glass exhibits absorption vs. dosage characteristics similar to those of phosphate glass described previously. Good linearity is obtained up to about 106 rad with less sensitive response at higher dosages. The primary advantage of this glass over the phosphate glass is its greater stability against fading as shown in Table 3.2. 3.11 Chemical Dosimeters Radiation dosage was measured almost exclusively by means of gas ionization chambers until the advent of nuclear reactors, and the largescale production of radioisoptopes. These developments led to a renewed interest in chemical techniques for measuring high-level dosage and also for measuring dosage in media other than air. The use of ionization chambers to measure radiation dosage is a very satisfactory procedure if the sample to be irradiated is air or some other gas. However, in a great many

100 9o0 i SPECIMEN UNHEATED so80 SPECIMEN AFTER HEAT TREATMENT 70 _. z HEATING PERIOD 13 MINUTES AT 1300C 10 _ 3O 0 2 4 6 8 10 12 14 16 18 20 22 24 1200 TIME AFTER CESSATION OF IRRADIATION (HOURS) Figure 3.19. Stabilization of Coloration of Silver-Activated Phosphate Glass by Thermal Acceleration of Fading. (48)

(5~) '.~$~am~so(I a~$~ssq~ awm~n-o-Ilr~f aq3-,o qcI'a::o3.q'0q' ~~~~~~~~~~~~~~-:j: i:-i:iiii-: a ---:: ':!i::::_:: i__ ~~ -i —'' -.. —:~i-:i ~:;l:-i —~::: ----':: ---:::..::::2::: —:i-~i~:ii /q: z ~:i:i-i-:-~ i i: i,!::t::;::~~~~~~~~~~~~:::i:::~~~~~~': ~~~~~~~~~~~~~~~~~~~~:.::.::::~:-:::::::::: ---: ~::j:.i..-~:,-i-i i-i-: I-:-i-~i: i -- I-i:l~-~: —:-~-': —: —:::: -: i/?::: J]-ri,:-:ii-; -:::- -:: -:::~~~~~~~~~~~~~~~~~~~~~:~~~~~.j'..:_-:i~-~i_ —i: -:::-:ii: -:::-::i:::-:: iiI~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~:::,l~i~-i~i:~:~c — -- —:.::::::.::::-i —~~ii~:-:::;::: -:-::: —:::-:_-~-:_::::j:::' J'?:'::i-: —.-.-:'.??i ii ': -: ~.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~::::_-_:::_i i~::~i-::::- i —i-:ii::

3-34 TABLE 3.2 FADING OF IRRADIATED COBALT GLASS AFTER STORAGE IN DARK AT ROOM TEMPERATURE(57) Fading (%) Normalized -at 1 hr. After Various Irradiation Doses Measuring Wavelength Storage Time 104 r 105 r 106 r 107 r 350 1 day No fading No fading No fading Too dense 1 week No fading No fading 7 to read 2months 3 5 9 400 1 day No fading No fading 4 Too dense 1 week No fading 2 10 to. read 2 months No fading 4 12 500 1 day No fading 2 7 8 1 week No fading 6 18 14 2 months No fading 12 23 23 studies, liquids and solids are irradiated and these have stopping powers quite different from gases. An additional requirement for accurate dosimetry is similarity in geometry, so, it is desirable to use a dosimeter which has the same stopping power as the absorber and which has a sufficiently flexible geometry to permit its incorporation in or on the absorber. The most practical chemical dosimeter should meet the following requirements: 1. The dosimeter should be readily prepar ed and stable in storage for long periods of time. 2. It should give a linear response to dose, independent of the energy spectrum of the radiation. 5.The product of the radiation-catalyzed reaction should be readily measurable by standard techniques of analysis and should be reasonably stable.

3-35 At the present no chemical system is known that rigorously meets all of the above specifications, although many systems have been developed that showpromise and are being steadily improved. Two reactions for use in dosimetry which have received considerable interest are: 1) the oxidation of ferrous sulfate to ferric sulfate; 2) the reduction of ceric sulfate to cerous sulfate.(58,59) Figures.21 and 3.22 give respectively the curves for the oxidation of ferrous ion and the reduction of ceric ion as a function of radiation dosage. The oxidation of the ferrous ion has received more interest and use than the reduction of the ceric ion and therefore will be described in detail..12 Ferrous-Ferric The ferrous-ferric reaction was originally studied by Fricke(6o63) and as a result the ferrous sulfate dosimeter is sometimes referred to as the "Fricke dosimeter.(59) The American Society for Testing Materials adopted this method in 1959 as a tentative standard for measuring the absorption of gamma radiation by chemical dosimetry.(64) The ferrous-ferric reaction determines the combined free-radical and molecular-product yields in an aqueous solution. In particular, thissystem is most reactive with hydrogen, hydroxyl, or hyperoxy radicals formed in the solution. The ferrous ion is oxidized to ferric ion by hydroxyl and hydroperoxyl radicals, as well as by molecular hydrogen peroxide: Fe~ + OH. Fe- OH (5.1) H + 02-. HO2 (5.2) Fe~ +HO02 -.Fe* +H02 33 Fe++ + H2O-~-Fe+++ + 0H-+ OH (5.4) The yield of the oxidation reaction, ferrous to ferric sulfate, has been found to be independent of the energy and of the dosage rate of the incident radiation over a wide range of the original ferrous ion concentration. However, it has been found that yields may be lowered by as much as 25 Iercent when dose rates exceeding 1,000 r/min are being measured.( 65,66) This method is sensitive to pH and oxygen concentration.

2100 _____ - 1800 ad w 1500 w a1200 w 03 ~0 -J' ~6 900 300 0 0 20 40 60 80 100 120 140 160 180 TOTAL DOSE (repxIO3) Figure 3.21. Micromoles of Ferrous Ion Oxidized as a Function of the Total Irradiation. (58)

33000__ __ ____- - j 27000 - __ _ - _ _ _ a. 21000. - _ __ W.j 15000- _ - 000.0 I 2 3 4 45 678 910 1112 131 5 TOTAL DOSE (rep xIO08) Figure 5.22. Micromoles of Ceric Ion Reduaced as a Function of the Total Irradiation. (58)

3-38 Ferric ion concentration varies linearly with dose up to about 50,000 rep. A linear relationship is observed at high doses as well, but with different slope as shown in Figure 3.21. Because of this discontinuity, presumably the result of the consumption of all the original oxygen, it is best to maintain a total exposure of less than 50,000 rep. Originally Weiss(58) recommended that the reagent solution used be prepaed with considerable care. The best reagent-grade ferrous amonium sulfate is'diluted in triple-distilled water acidified with sufficient reagenb-grade sulfuric acid to prevent hydrolysis. Later, Weiss stated that small quantities of sodium chloride (0.15 gm NaCl/gm FeSO4 ) inhibits organic impurity effects so that solutions may be made from ordinary material. The ferrous salt is added to the 0.8N H2SO4 but the exact concetration of ferrous ion need not be determined as the analysis depends upon the determination of the concentration of ferric ion. To analyze for ferric ion concentration it is necessary to prepae a standad reference ferric ion solution. Weiss recommends commencing with a solution in which the ferric ion concentration is about 0.001 - 0.00 M. This solution may be completely reduced by means of a Jones reductor and titrated with standard permanganate. A known amount of this reduced solution may then be quantitatively oxidized with peroxide and diluted to a known volume to prepare various concentrations of standard reference ferric ion solutions.; These standard ferric ion solutions are then used to prepare a curve of optical density vs. concentration. The optical density is conveniently determined by means of a Beckman spectrophotometer. The maximum absorption for ferric ion occurs at about 304 mjj. However, this'absorption peak is rather broad and the results are not very sensitive to the wavelength. Weiss used a setting of 505 M~t with a slit width of 0.5 mmn, quartz cells and a hydrogen lamp. In the analysis, a nonirradiated sample-of the ferrous sulfate solution is used as a blank and the difference between the optical density of the irradiated solution and the blank is determined. This difference in optical density permits the calculation of the oxidation of ferrous ions by means of the calibration curve of optical density vs. micromoles of ferric ion. This determination permits the total dose to be estimated by means of Figure 5.21. By means of a conversion constant, the calibration curve may be combined with Figure 5.21 to give a curve which reads directly in kilorep dosage as shown in Figure 5.25. Any laboratory using this method should prepare its own optical density calibration survey as different instruments will give slightly dif

OPTICAL DENSITY ON BECKMAN MODEL DU METER 0 '__ __ 0 0 in a) ) a 0 - z m:I 0 0 Figure 3.23. Calibration and Conversion Curve for Beckman Model DU Meter in the Fission Products Laboratory, The University of Michigan.

3-4o the Fission Products Laboratory and using a conversion constant of 15.6 micromoles er liter per kilorep.(67) A further precaution is that for best results dosage to dosimeters should be limited to less than 50,000 rep. The usual procedure in the Fission Products Laboratory is to limit t xposure time of the dosimeter so that the estimated dose is about 0,000 rep so as to stay below the break in the curve. Although this reaction is extensively used, it is not especially good for high-intensity sources because of the high G value and the fact that the iron oxidation is dependent upon the oxygen concentration. The G value stated above for oxidation of ferrous sulfate holds for C6 gamma radiation and electrons. However, the G value for this dosimeter varies for radiation of different quality and energy(68) and hence, a different conversion factor should be used. For a given radiation exposure, ferric ion yield has been found to depend upon the diameter of the container used. This effect is resumably due to the increased production of secondary electrons in the vessel walls as compared to the production in the solution. Valid absorption measurements can best be obtained if the dosimeter vessel approximates the vessel to be used for irradiation. If this is not possible, the diameter of the dosimeter should not be less than 8 nm.(5859) Ad/ditional information on the ferrous-ferric reaction and its use in dosimetry is given in References 69-76. 5.15 Ferrous Sulfate-Cupric Sulfate Thferrous sulfate-cupric sulfate dosimeter is fairly new in the field. It eliminates some of the difficulties experienced with the Fricke dosimeter. The reaction consists of the oxidation of the' ferrous ion by hydrogen peroxide formed in the gamma-ray 'Thotspot" and is not due to oxidation by free radicals. The addition of cupi ion to the ferrous-ferric system reduces the G value from 15.6 to.o.66.(77) The reason proposed for the decrease in ferric ion yield is that the cupric ion reacts with the hydrogen atoms and hydroperoxy radicals as follows: Cu++ + H *-Cu+ + H+ (355) Cu~ + H02 -7.- Cu + H + 02 (5.6)

3-41 reduction in the ferric ion yield is effected by the reaction: Fe+++ + Cu+_- Fe++ + Cu++ (7) Thus, in the presence of cupric ion, the ferrous ion is oxidized oniy by molecular hydrogen peroxide, two ferrous ions being oxidized per molecule of hydrogen peroxide [cf. Equations (.1) and (.4)]. This system has been proposed as a chemical dosimeter(77,78) offering the advantages of oxygen independence and applicability over a wide range of dosage, as indicated in Figure 3.24. The ferrous-cupric dosimeter has been employed in measuring gama dosage from a cobalt-60 source and alpha-particle dosage from the boron(n) lithium-7 nuclear reaction. It has been found that the hydrogen peroxide yields are higher for the heavy alpha-particles than for the gama,photons..1 Ceric Sulfate Ceric ions are reduced to cerous ions by the ionizationproducts in the aqueous solution. The yield is linear with dose, indepe ndent of concentration over a wide'range, energy independent from 100 kv to 2 Mev, and independent of dose rate. No difficulty is encountered with the consumption of oxygen in a closed system, as in the ferrous-sulfate reaction. The G value is about 2.4.1 molecules reacting per 100 ev energy absorbed.(58'70) The discontinuity in the cerous ion dose curve occurs above 10 million rep (see Figure 5.22), indicating that up to this total dosage the reaction is independent of the oxygen concentration. This makes much higher total dose measurements possible than with the ferrous-ferric ion system. Another advantage that the ceric-cerous system offers is that the yield is not diminished at high dose rates, as is observed with the ferrous-ferric system. However, trace amounts of organic compounds lead to erratic results, and also at low ionization energies the system shows a dependence on the ionization-energy'spectrum. Although ferrous sulfate titration may be used,.the most convenient method of analyzing the irradiated ceric -cerous solution is by means of a spectrophotometer. In this technique the absorption by ceric ion at 510 - 520 mp. is measu-red.(65) This technique gives an accu.racy of 5 percent in determining dosage. Although this reaction appears quite suitable for high intensity sources it has not been investigated as thoroughly as the more common ferrous sulfate reaction. It has been reported that considerable care must be used to obtain reproducible ---4- - reuls.Inernty however n,the r acracy ofnr, -

600 500 400 / + + 300 IL %M OO# z 1200 t Fricke dosimeter I00 0o Ferrous-cupric dosimeter 0 0 50 100 15020 200 250 -20 ev /LITER x 10 Figure 3.24. Comparison of Response Curves for Fricke and Ferrous-Cupric Dosimeters. (77,78)

3-43 the determination of the cerous content should be much better than the similar ion determination with the ferrous sulfate dosimeter since the change in optical density per unit change in ion concentration is greater.(79) 3.15 Chlorinated Hydrocarbons Ionizing radiation may decompose halogenated hydrocarbons to products which can be detected in aqueous solutindicators. This princi le has been used in the preparation of chemical dosimeters.(2,01,80) -In the absence of inhibiting agents these reactions procede by a chain-reaction mechanism which is very sensitive to temperature and impurities. By introducing resorcinol or certain other alcohols (e.g., ethyl, hexyl and decyl) to the system, the long chain reaction is inhibited and: the attendant disadvantages are eliminated.(81,82) Thus, the alcohol acts as a stabilizing agent. 'This type of dosimeter may be either single-phase or two-phase.(83) In the single-phase system a small quantity of the halogenated hydrocarbon and a dye ae dissolved in water to form the dosimeter directly. In the twophasey (84) the halogenated hydrocarbon forms a sepaatephase. Both chloroform and tetrachloroethylene are used as reactants, although the latter is-Preferred because of its greater sensitivity to radiation and its greater temperature stability. The indicator employed most commonly is Bromreso puple.The echiques of preparation aalysis, calibration, and se f ths tpe of dosimeter are described in detail in the literature.2 Taplin described(85) three types of chemical systems, prepared from chlorinated hydrocarbons stabilized with resorcinol and using aqueous pH indicator dyes. One system consists of a chlorinated hydrocarbon overlayered with a pH indicator dye. The second system is prepared by saturating an aqueous pH indicator dye with relatively small amounts of a chlorinated hydrocarbon., Both of these systems can be adjusted to respond equally to gamma radiation. However, they differ by about a factor of 5 in their response to fast-neutron irradiation. By using both systems simultaneously in a mixed neutron-gamma field it'is possible to estimate the separate dosage from each type of radiation. The third type of system described by Taplin(85) uses tetrachloroethylene, which is gamma-ray sensitive and devoid of hydrogen. Dosimeters using tetrachloroethylene are not affected by fast neutrons and therefore can be used to measure gamma radiation in a mixed field of gamma radiation and fast neutrons. Hilsenrod )and Andrews et al.() have studied irradiation of

3-44 University of Michigan indicate that the pH of a 0.2 M choral hydrate solution vaies in a linear manner with dose and the change in pH can be used to measure doses from 100 to 1000 rad..16 Gaseous Nitrous Oxide Donds(88) and more recently Harteck and Dondes(89) describe a Simple highlevel dosimeter using nitrous oxide gas forthe detection of gamma, or thermal neutron radiations. The advanta nitrous oxide are its ease of purification, indefinite shelf life, and usefulness in temperature ranges of -80~C to 200~C. Also, the decomposition products produced by radiation consist of oxygen, nitrogen and nitrogen dioxide, which do not react with each other and which may be easily measured by vacuum techniques at any convenient time after irradiation. The procedure used for ionizing-radiation dosimetry consists of filling a 20 cc quartze vessel with purified nitrous oxide at 500 mm Hg pressure at C, sealing and exposing the filled vessel to radiation and then analysing the contents. The filling conditions of 500 mm at C correspond to one atmosphere of air in defining the roentgen.(89) For dosages up to x 107 r the dosage is determined from the analysis for N2 and 02. For dosages between 5 x 107 and 5 x 109 r., the dosage may be determined in the same manner or more easily by colorimetric analysis for nitrogen dioxide without opening the dosimeter. The dosimeter approaches saturation above 5 x 109 r and loses accuracy at higher doses. The reactions involved are given(89) as follows: N20* N2 + 02 80% (3.8) N20 AN20~ + e N +NO 20% (5.9) 2N0 + 02 - 2N02 (5.10) 2N0 (5.11) N02 +N — N20 +0 (5.12) N2+ 02 (5.15) N20O+ + NO ---— >- N2 0 + N0+ (5.14) NO +- e > N + 0 or NO + hv (5.15) The primary yield (80 percent) from the irradiation of N20 is N2 + 02 as indicated-v byEquaion4n (5.8). Thisnl mynr resui-lt ihrfo+h

3-45 result of interaction of the radiation and N20. A smaller yield of N + NO (20percent) is produced by such interactions. These intermediate products enter into additional reactions indicated by Equation (10) through (315) to product NO2 (Equation 3.10) and more N2 + 02 Equation ). Harteck and Dondes state the decomposition of nitrous oxide is linear with dosage up to 107 r and almost linear up to 8 r as shown by the calibration curve in Figure 3.25. If the dosimeter is to be used to measure thermal neutrons in a reactor, 5 mgf U-235 oxide is inserted in the dosimeter as a owder before the addition of the gas. With reactor irradiation the neutroninduced fission results in about a 60 fold increase in the decomposition products over that obtained by reactor irradiation without -25 oxide powder in the dosimeter. E.periments were conducted using pressures in the dosimeter from less than one atmosphere to the critical pressure of N20. Temperatures were vaied from room temperature to 150~C. Irradiations were conducted with cobalt-6, fission products, and in a reactor. The intensity of irradiation varied over a range of 4 decades. All results indicated(89) that in the linear region up to 107 r and almost up to 10 r, the nitrous oxide deconroses independently of pressure, temperature, radiation intensity, dose rate, and decom~position pout rsni h ai given by Equation (5.16) N2: 02' NO2 =1:0.14:0. 48 (5.16) 5.17 Other CemicalSystems Chemical dosimetry with various other systems have been studied and some are indicated below. A review of chemical dosimetry methods was made by Harmer(90) in 1959. 1) Various dye solutions are suggested(9l-93) of which aqueous methylene blue could be used up to a dose of 6 x 106. 2) Stein and Day have described a system of benzene and sodium benzoate in water.(94'95) Armstrong has described use of calcium benzoate() in water solution for use in low dose' measurements. (96) 5) ]Draganic has described use of oxalic acid as a dosimeter

I00~~~~~~6 U) * BNL eactor 235 0.01 I~~~~~~~~~~~~~n pie Figu~re 5.25. Calibration Curve for N20 Dosimeter. (89) (o~1gsae

3-47 Other hemical methods include the production of iodine from iodides, and the decoloration of DPPH in organic solvents. Some investigations havte been carried out with certain dye systems combined with gelatin or agar(98l1l) to form a gel. This type of ticulaly usel1 to measure dose as a function of depth for electrons. Additional information on chemical dosimeters is given in References 102 -123. 3.18 Calorimetric Dosimeters Calorimetric dosimetry techniques are based on the fact that part or all of the energy of the ionizing radiation absorbed by matter is converted into heat energy. By measuring the rate of heat generation inthe substanc, the dose rate can be determined. If the substance is not chemically changed by the incident radiation, all the radiation that is absorbed is quantitatively converted to heat. If the substance is chemically altered by the radiation, part of the radiation goes to effect the change and the remainder is converted to heat. The very small magnitude of the heat energy produced in matter by ionizing radiation requires extremely sensitive techniques and equipment for calorimetric dosimetry, although the basic principles involved are the same as those of conventional calorimetry. As an illustration consider the thermal value of one roentgen. (24 1 roentgen 93 ergs/grsm. (3.17).1 calorie =4.18 x 107 ergs/gras. (3.18) 1r=2. 22 x 10- calories. (3.19) Thus -a dose of 1 million r corresponds to a temperature rise of only about 2.20C in a water-equivalent material. Most experimentalgamma sources currently in use produce dose rates that do not exceed 1 megare~pper hour; therefore, a very sensitive calorimeter is necessary. The calorimeter consists of a thermally insulated substance used to absorb part or all of the ionizing radiation and temperaturesensitive devices to measure the temperature of the material as a function of time. Knowing the heat content of the material as a function of temperature,, the heat energy absorbed per unit time can then be determined. Two criteria are used in selecting the absorbing material for maximum sensitivity. First, it must, have a high linear absorption coefficient, that is, a given mass should be able to absorb a large fraction of the incident radiation.

3-48 gives the linea absorption coefficients for water, aluminum iron and lead together with their densities: TABLE 3.3 LIEAR ABSORPTION COEFFICIENTS AND DENSITIES(125 LINEAR ABSORPTION COEFFICIENTS FOR Energy, (Mev) Water Aluminum Iron Lead 0.5 0.090 0.23 o.63 1.7 1.0.o67.16.44 77 1.5.057.14.40 57 2.0.o4.8.12.33.51 5.0.038.09.30.47 Density, gm/cc, 200~C.998 2.699 7.86 115 Of the four materials listed in Table 3.3, lead is the best absorber for calorimetry. The most accurate methods of measuring temperature changes in the absorber are resistance thermometry and thermoelectric thermometry. In applying the techniques of resistance thermometry., use' is made of the fact that the resistances of electrical conductors change with temperature. Geneall, mtal hae apositv teperature coefficient of resistance, that is, an increase in the resistor temperature causes an increase in resistance. Nonmetallic conductors may exhib it a negative temperature coefficient. Some transition metal (nickel', cobalt, manganese) oxides have high negative temperature coefficients -(about 5.9 percent per degree C) and are often used in resistance thermometry. In thermoelectric thermometry the thermoelectric effect is'employed. If two wires of dissimilar metals are laid side by side and connected at the ends;' a temperature difference between the two ends will cause a difference of potential to be developed. The thermocouple so formed may be. used to measure an unknown temperature with one junction by maintaining the other junction at a reference temperature, usually the ice-point. A number of calorimeter designs have been developed for use as dosimeters. Laughlin et al. have described a calorimeter which uses lead as an absorber.(l5l~ Lazo et al. has described a calorimeter which usesc- watr.r as anc,-nabsorb-,- -Ter i- n a simpleI pyre bul cnta.in i ngaqtheT(rmocouplefl27)

3-49 With this e the error was kept at + 2%. The same grou later described calorimeter with an absorber made with polyethylene and free carbon. A graphite sphere calorimeter is under development at the US National Bureau of Standards and is described by Hart et al79) A graphite sphere suspended on polystyrene pegs inside a graphite shell which was heated toprovide an adiabatic shield was used. This was exposed to a dose rate of approximately 5 x 105 r/hr from the NBS 2000 curie cobalt source. An error of less than 1% was claimed. Taimuty has described a liquid calorimeter for calibration of ceric sulfate dosimeter.(128) He had also described a solid calorimeter for measurement of electron-beam intensity.(129) The latter could be used rption in nonconducting solids by making mea ments with and without the solid interposed in an electron beam in front of the calorimeter. Figures 3.26 and 3.27 show one type of calorimetric absorber and measuring-chamber geometry. (126) The thermistor is used to measure the temperature of the lead cylinder. The calibration heater is used to determine the heat capacity of the absorber in the absence of radiation. The purpose of the radiation shield is to reduce the heat transfer by radiation from the absorber to the measuring-chamber wall. The absorber and radiation shield are suspended by nylon threads in the measuring chamber, which is under vacuum. The vacuum reduces heat transfer by convection from the absorber, and the nylon threads reduce the heat transfer by conduction. The nylon also has a relatively good resistance to radiation damage. The absorber assembly outer surface, radiation shield, and vacuum-chamber inner.surface are all chromium plated to reduce further the heat transfer by radiation (gold is an even better surface for low emissivity). The measuring chamber is immersed in a water bath, whose temperature is controlled thermostatically. 5.19 Lumin iscence Degradation This method of measuring radiation dose is relatively in an early stage of development but offers a promise for measuring large doses. The method is based on the decrease in luminiscence of many organic ~1u~no~p1oi after exposure to high energy radiation. An "exciton" theory(1112 suggests that radiation damage consists in introduction of impurities which cause luminiscence degradation. In a nonirradiated crystal the excited energy o-riginally i-mparted byv radiation is transferred between molecules by resonance

LEAD ADSORBER TUNGSTENBRSSHL WI-RES HEATERr __ _ _ _ _ THER STOR LJl IMBE__ __ __ __ _ ___DOD IMID E INETOOVm.1INWOS METAL C HE RMBISTON FRODn IMBEteD

BAS VACUUM CHAMBER ALUMINUM RADIATION SHIEL SILVER PLATED MASONITEo BOX MERCURY CENTR I FUGAL 0 THERMO PUMP THERM ISTOR RGULTO 0HEATERS HEATERH WTER COLE ALMNMWATER BATH TUBES ~~~~WINDOWS Figure 5.27. Horizontal Cross Section Showing Calorimeter Construction. (126)

length outside of the fluorescence band of the pure material. This effect can be used for dose measurements by observing the post-irradiation luminescence at a frequency in the visible or near-ultraviolet spectrum in response to a higher energy excitation. Attix has described systems using wafers of anthracene and pquarterphenyl which are excited at 3650 A and read through a filter with a pass band centered at 4420 o.(133,134) He described a certain amount of recovery of luminescence which is accelerated by heat treatment for 1 hr at 1000C and stabilized the readings. Barr studied the destruction of the fluorescence of quinine in acid solution by 250 kvp x-rays.(135) Quinine concentrations of 10-5 to 10-8 moles/litre were used to measure doses from 5 to 2 x 104 rads. The decrease of scintillation efficiencies of crystal, plastic, and liquid scintillators could possibly be used for dose measurements. Rozman reported that damage to plastic scintillators is independent of dose rate but dependent upon the nature of radiation. (136) 3.20 Microbial Monitors Microbial monitoring as a method of dosimetry could be applied to specialized applications like radiation sterilization. Various techniques have been described. (137,138) However the choice of microorganism for monitoring should be made carefully since there are marked differences in radiation resistance among the various species. 3.21 Summary The various dosimetry systems proposed have been described. The choice of any one depends upon various factors in a specific situation. A wide number of dosimeters are now available for doses of of~~~ thehdo nlsswt epc t ieadcs.Gaso lsi

3-53 dosimeters do present problems of instability. However, theypresent the advantages of convenience in storage, extended life and rapid and easy measurement. Costs of equipment may be considerable in certain methods iof dosimetry, but the added expense may be worthwhile because of convenience and precision of the measurement. For doses larger than those mentioned above one has only a limited choice. Luminiscence degradation and nitrous oxide dosimetry may be used. Schall has compiled a table from questionnaires sent to differen eperts on the evaluation of dosimetry 'Methods. This is reproduced as Table 3`),`, (140) with-slight modifications. A table of characteristics of various secondary standard dosimeters has been compiled by Hart. (79)

Table 3.4 Chemical Dosimeter Systems (Courtesy Nucleonics, McGraw-Hill Publishing Co.) (90g Materials Time for De0ay Reproducibility Equipment t ea t Fraction of NO. Dsimety sysem Toal-doe rane (ras) Doe-rat range (rad/0) cost (dollars (min) toh 2+)ge that 20+. 000022t 2y~~~~~t22 Tot~~~t-0o22 o~~o~~ (0002) %(dolla2,( per I 20 too (musgt (can is permanent Advanttages Disdvatages measurement s b) b I Ferous-ferric20+0(Fik)4xL to+ - 4 04 02 00t o3-t106 2,500 0.00 2 20 1000 0 toO5 too high acoura+yjdirectanalysi nedo operat00or experie withoit further treatmet;so+0- need foo(sre in) 02adiato N~~~~~~~~~~~~Hr)tonc. ocupysaeadvl; of 22250, 2 Ceric-,erous 5 x I4 -2 xt08 +0 LoO -LOS tO0 or 04.05 30 200 t,000 o LO3 too simplcty; accuracy need fo titraton t tigb (osity 2,000 dsesandrates 3 Ferr~ous.uopric4 1o0- 107 05-to xL o~o- 8o oo 2,000 0.ot 5 00 200 0 tO03tO20022~~200.2tt00t22~tt~ (Swope) 100 extends Rageprou iyitit D~olt-raon-s ae r tane adr (s~~~~~~~~~~~~~~~~~frin) sysatem g eros-Disadvantaes; oj (del~,s)~~~~~~~~~~~~~~~~~~~~~~~~~~~~~2...2 000 0mut,pn me~~~~~~~~~~~~~~~~~~~~~~~~~~~~asueet hig does)b 4 Herogeusted-feyrocaric o- o0 -to7 e5 4 x toO - oxO 2,0000 acidimetric~~~~~~~~~~~~ dye 1031 - 5 10 3,00* 0.0 OD2 100 s 0 ighacurc; dinsecanasitiiyt etos; need ofoeaor eperodcionce;v (S~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ~ ~~~~~~~~~~~~ithoutfurhe treangen; s;eoflvauato need for glass intirdatione (00s002) tincnocp8saeadvlm a0er plastic v0i0) o Nitrous oxi0, t+5 4 x 10 S to7 -o9 500-t,000 0.250 or 55.30 500 2,500+ 0 t0O 100 tsmplicity range - 0-20; acc a need for titaou an0t-i hui (Harteck & Dondes) dec~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~Oom~position productsdon't need to break seals in va (Hart~ 00+0,2). 0+t22,0t 002 doese and rates 6 Methytebtue t104 - o07 + 25 104 - 2 x 108 250 0.01 - 0.00 t5 6o t 0o o tO 100 ex se ansmpitty for rou0ie0need+for0 standardization off (+00oth) ferri rane 0oluion les tbl0ta 7 Gasevoloutio 2 jot05 -2 t07 + -5 t4- t x 0 -O10 200 <0.00 30 >t tOO fot co0t0000+2 needfor individual tto. cidimetricye (Sigoloff) ran+t ~ge; 2,se ofev aluation need tforg ss c ontainercior al rtt iations incd toing nutrons 2ut 8 Starch-iodine 4 x to3 - to5 300 0.25 0 05 60 60 to103 too0 simplic2ty-.of clormtr+ meas- aenerg d2pedence; 20astarc via N i t rousGoxide 105 - x 108 + 5 107 - 109 500-1,000 O.25* ora30 5O0 2,5007 0 temperaturement; little -0~-00~need for tteacmustbeslublem need lit atr; decomposition to p+rifi iont; r2 2- 02e2 0+ or color g 0 vto justment by vary252 io0i+e 0022+2-20+0v202+0 (btt 00t02 (Harteck & Dondes ) i n t e r a c t; w i de dose and rate~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~cocntatonbemae o (000gr) y(0 0 p0p0) 104 _ -07 +2 2 05x L0 0.00 0.3 5 20 0 >5,000 000 speed; cos0; simplicity; 020k 0200 of +o02r2i02 sandadizabti (GoNbith) (visua evaluatchon)dy instrmna evaluation) 4 LO exosr 00 Dyedplsi ( y 2 x 104 -42 x107 2O5 003 - 2 x 31000 0.001 5 2* 0 ) 1 000 con00+ i+0+e; cost; simplicity 02ria0+0ndu0 accgas-yoiec (0art~ surc monitoig respone to2 tion ytmfreahmaue to 0arke+0n8 of plastis t -ioie 5 x 10o - 2 0 0000 600-1,200 0.00 10 simplicity >f c o lo0 i mety; ost; applicabil- 0220 fo0 standar2 c+0or 0,0 (0202000) 2ty 22 202+20 2220+20000 22t0~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~t o isalealato mn 02 Depolymerization 0 000o3 -+00 8 + 5200104.- 5000lo7 200 0.02 3 6+ 200 0 103 1000 2002 2ang (40de0ades2f0r0+e liitned fouratoe y; need for (Gerna) doiometoeprifcto; 6 a g dd ieclade for prtr conrl or gorowthno Justsem nd tandaryin reading timse-euvln bti a 13 Photopymriaon- + (20 20ene107 +2-5 5 x LO3 - 2 x L0050 0.0.3 0 0 2 00 go-0n-go2sy0t2; 0222rn2; decrease of5sharpnes 0200f 00022022) (visalevauaio) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ polesersd;o~k cost; sipiiy aklack of nedcommrca auxil- atbighsloe;rqitym or nequpet ofr processing aforterod 10 4 14- 0 10 0022las22s 0+2222 2 S n x 0 - 4 x 5+ 23 -.5 x 1 00 250 -,350 0.5 00 50 05>5 2 tO5 00-too convenience; c o e ss; simp lidityef 0 aat2in t t at d0 (Henley2+) 02ere000 2atches 2002082, 00220l02 a 22+s2t(00-00, undo~~~~~~~~sent 15 Lumi2e2+enedegrad0tio x o0 -.000 +_ 5 5 x to3 - 5 x o0 7 200-300 0.05 0.00 0.5 00 50 0> >oo 4de range4(dfor one'. Depolymerization (Fan g ) l i m i t e d a c c u r a c y; n e e d f o r tem-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ 9t malns; old;loger nedfo ba teamet i (Schulman) st~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~~~~~~~~~~~~~~~oragee; doecde for pnoe r a isto avoidtrol orecovryrecto (sytem2 Of -eapa, readingy 222il 2002 002mi+202; 0002 range; 2000 0 2o2rereati; osimplity 16 000+000lator damag2 (q+0+002 2n2 5 x jo3 lo4+2 5 x 0o3 - 5 x 107 0,000 0 0 20 000 0 yt3 000 pere2; cotvenience; reaproduci 0 0 garipn 222es se acid) (02Bar) biligt; se+ui02m00y 17 Glass fluoresenin xrjo3 lo6 +2 5+003.5+000 250 0.30 _ 05 6+ + to5* 50.95t 202l0000 a0ura1y;o+t03 nee0d8for0a0batio;0limt (Krdhu lmair) ruggdefniessoae dosedo range(0-,00r L200 - 0000202; 0 50 +5 tO Pho002000200 cells s+ t o3-LO to9s + to3.0-00 -Oso00, Ol t t. t simplicit; s eed (Moody)Schulack of the-p, response 09 Cerenkov.photovoltaic 2200002 02t 6+0 06-2.2+007 02 LOo- 6+xoo7 50 o 1 * 0 0 simplicity; cost;0lack of need for radiatio 22e2ergy 0 0022 chemcal; 00id20e0 20080an00ge0;20 2 device t~~~~~~~~~~~~~~~~~~~~~~ ~ ~~~~~~~~~~~~~~~~aimedlay oaithmceog to Prodc eread; sipict (Th0ms) 4o -6 0i5 0 )titor elecons in + 20 020miu0-2+0l2002 (220, device) 00se 0 x 0O3 - 6 + o0 +2 5 x tO0 - tO 6x0 0.50 - 2 0.2 3 05 0.52 4 +o04 (litte)n 0c,00 (0000a2002) 0ici00 02s0 feasibility (doeidevBice) to 20 Far0day5c2ge2, 0+22, 0020222 0+5.- 2 x.000 02.5 x 000- 0.5 x00 0 200 0 - 0 0 0 0 --- 0 e00ctricl00atur; applicabil- 2200022 0200202020,00+22 0- simplicity;accuracy osf nee d fo r c alibrationo8; l iginaltdesin of 22 Ionizaion cambers5 x L s~o-t io 5+003. -5+x0000 < 500 — 2-2 --- 0di+0ect raig;availability0of (02,02002000) recyc0008; 0000 020d 20e02 l02e 23 020050220rx 1+o0.s- xo7 02 2, 104 -.2+0 109 2,500 -— 06 —20-0 — +02002200 (Tai2+05) 00ly; 02,0 202 s0000e0 per20 0 20 Microbial monitors 4 x o3 -ox o6 020 5+003O.-5+ 0000 200 or 0 05 020 6+0 1o4 +2 1000 directmeasuement f steili- ned0fo 22 27 days00to0obtain+00 2,500t za0200; applicability for 22- need f+2 bacteriologi+al 022+00 pruggessadfo ness do steran18 Photovoltaic cells + 5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~00220 020 020000 t~~~~~~~~pecti-~~~~~~~~~~~~~~~~ ~~OptO*ee mtOd Me phtn ca be exiplty;nsped bylactrockil.tie ea ramn Ih ag of total-dose haspnotb nvsti ~~~~~~~~~~~~~~~~~~ Haloge202te0l0ydrocarbon-600dimetri0. 50 2000 3-0+. 0,20 2000er 100090) 02 pe222+2+0 2000002,2es ofLo - 07 220 ' 82020; 4x L 00 -+22 12202 tx 108 +as/e to 0 0+2+ +O;22 20 0t Oyo80 02 +0 022000+Dy0220+;>0222220 02 82208 00500+2 0+22 020,0;00 2tsg20 030+00. +a0ene22y 222sors2to determi 02 ty +2 *l2 i u d +22uc22+00Dy 02 222+ g*times 2su002.. darenig:0*dela0 bew- 0+en 2+2 0002 5,0+ 2;022; is t00, 2000 +0 22p222+0 damage; treproduc.il 00+iza2i+n 00a2+r ti,2 020200 0 0+00.;.0....222. automati0 (eading 2002220222(r 020+02 ag ing 0220+0, o+ pa 2- 8022 0 22 22+008eif Y 222+000 f itis 2 %; 200002t0 20urc 22+00 0020 2+0 002 0+22 Nitrous 222220; +22,200.....P 0+2 +0000aotr polyeste2220 +220+2g000220;-620x 2(2t c0+22pgl 5o 02m 202 fadi 02+200 0s - 2 2 0 0 +000 o0022 byc2h22000202g00e+00 020b0202+22,.0220222det0ermined;222 2+0 0200) 02 2202 022202.ay 2000+02... 22.2.+0+2.0bi montors + 2220 +22z 2v00+000+; 0222cial apparatus could -no eqpuipm22 e+020 n2+t2222202+0 2222022e 22220220 20d 3,500i 2+0red; oA 202 00+020os02s25.02+ 0202+00-2+0200,f w00h;e-2i2+0+8t22o2002;n 0222 bacte 02020l 22+0 NUCLEONICS 02, No. 0, 40 (1950)] reused a20er a20e202+g 2+2 5 00 a0 4oo~o +ut pre00si0+ 02 s2000 witins 5% (f tonec 2220s 0 wate2, essentially

CHAPTER 3 REFERENCES 1. Gurney, R. W. and Mott, N. F. "The Theory of Photolysis of AgBr and the Photographic Latent Image." Proc. Roy. Soc., A, 164, (1938) 151. 2. Hine, G. J. and Brownell, G. L. Radiation Dosimetry. Academic Press, Inc., N. Y., 1956. 3. Ehrich, M. and Fitch, S. H. "Photographic X- and Gamma-Ray Dosimetry." Nucleonics, 9, No. 3, (1951) 5. 4. Tracerlog No. 62, "Neutron Film Badge Service." Tracerlab, Inc., Boston, Mass., Sept., 1954. 5. Anonymous, "Commercial Film-Badge Services." Nucleonics, 13, No. 2, 1955. 6. Beiser, A. "Nuclear Emulsion Technique." Rev. Mod. Phys., 24, (1952) 273. 7. Cheka, J. S. Neutron Monitoring by Means of Nuclear Track Film. ORNL-547, Oak Ridge National Laboratory, Oak Ridge, Tenn., 1950. 8. Cheka, J. S. "Fast Neutron Film Dosimeter." Phys. Rev., 90, (1953) 353. 9. Cheka, J. S. "Recent Developments in Film Monitoring of Fast Neutrons." Nucleonics, 12, No. 6, (1954) 40. 10. Cobb, J. and Solomon, A. K. "The Detection of Beta Radiation by Photographic Film." Rev. Sci. Instrum., 19, (1948) 441. 11. Deal, L. J., Robertson, J. H].. and Day, F. H. "Roentgen-Ray Calibration of Photographic Film Exposure Meter." Amer. J. Roentgenol., 59, (1948) 731. 12. Demers, P., Lapalme, J. and Thovenin, J. "Fading of the Latent Image Formed by Charged Particles." Canad. J. Phys., 31, (1953) 295. 13. Dudley, R. A. The Measurement of Beta Radiation Dosage with Photographic Emulsions. Ph.D. Thesis, The Massachusetts Institute of Technology, 14. Dudley, R. A. "Photographic Detection and Dosimetry of Beta Rays." Nucleonics, 12, No. 5, (1954) 24. 3-55

3-56 15. Ehrlich, M. Photographic Dosimetry of X- and Gamma Rays. NBS Handbook 57, U.S. Gov. Printing Office, Wash., D.C., 1954. 16. Goldschmidt-Clermont, Y. "Photographic Emulsions." Ann. Rev. Nuclear Sci., 3, (1953) 141. 17. Greenberg, L. H. and Haslam, R. N. H. "On the Shrinkage Factor of Nuclear Emulsions." Canad. J. Phys., 31, (1953) 1115. 18. Greening, J. R. "The Photographic Action of X-Rays." Proc. Phys. Soc., Lond., B, 64, (1951) 977. 19. Hoerlin, H. Development of a Wavelength Independent Radiation Monitoring Film. ANL-5168, U.S. Atomic Energy Commission, Wash., D.C., 1953. 20. Jetter, E. S. and Blatz, H. "Film Measurement of Beta Radiation Dose. Nucleonics, 10, No. 10, (1952) 43. 21. McLaughlin, W. L. and Ehrlich, M. "Film Badge Dosimetry, How Much Fading Occurs?" Nucleonics, 12, No. 10, (1954) 34. 22. Mees, C. E. K. The Theory of the Photographic Process. Macmillan, Co., Inc., N. Y., 1954. 23. Rosen, L. "Nuclear Emulsion Techniques for the Measurement of Neutron Energy Spectra." Nucleonics, 11, No. 7, (1953) 32. 24. Rosen, L. "Nuclear Emulsion Techniques for the Measurement of Neutron Energy Spectra." Nucleonics, 11, No. 8, (1953) 38. 25. Simons, H. A. B. "Use of Nuclear Emulsions for Fast Neutron Dosimetry." Nature, Lond., 168, (1951) 835. 26. Titterton, E. W. "Slow Neutron Monitoring with Boron and Lithium LoadNuclear Emulsions." Nature, Lond., 163, (1949) 990. 27. Titterton, E. W. and Hall, M. E. "Neutron Dose Determination by the Photographic Plate Method." Brit. J. Radiol., 23, (1950) 465. 28. Tochilin, E. and Golden, R. "Film Measurement of Beta-Ray Depth Dose." Nucleonics, 11, No. 8, (1953) 26. 29. Yagoda, H. Radioactive Measurements with Nuclear Emulsions. John Wiley and Sons, Inc., N. Y., 1949.

3-57 30. Army Chemical Corps, Technical Command, Symposium No. IV: Chemistry and Physics of Radiation Dosimetry, Part I, Army Chemical Center, Maryland, Sept., 1950. 31. Army Chemical Corps, Technical Command, Symposium No. IV: Chemistry and Physics of Radiation Dosimetry, Part II, Army Chemical Center, Maryland, Sept., 1950. 32. "The Effect of Atomic Weapons." Chapter 7, U.S. Atomic Energy Commission Wash., D.C., (1950) 238. 33. Miller, A. Gamma-Ray Dosimetry with Polyvinyl-Chloride Films. B.S. Thesis, Department of Chemistry, Brooklyn Polytechnic Inst., Brooklyn, N.Y., 1951. 34. Henley,E. J. and Miller, A. "Gamma-Ray Dosimetry with PolyvinylChloride Films." Nucleonics, 9, No. 6, (1951) 62. 35. Brownell, L. E., et al. Utilization of the Gross Fission Products. Progress Report 5 (COO-196), Engng Res. Inst., Proj. M943, The University of Michigan, Ann Arbor, Mich., 175, Sept., 1953. 36. Henley, E. J. "Gamma Ray Dosimetry with Cellophane-Dye Systems." Nucleonics, 12, No. 9, (1954) 62. 37. Henley, E. J. and Richman, D. "Cellophane-Dye Dosimeter for 105 to 107 Roentgen Range." Analyt. Chem., 28, (1956) 1580. 38. Row, F. M. "Color Index.' Society of Dyers and Colourists, Index, (1924) 518. 39. Personal Communication to L. E. Brownell from Mr. C. L. Blair, Sales Development and Technical Services, I. E. DuPont De Nemours and Company, Inc., 1956. 40. Brownell, L. E., Meinke, W. W., Nehemias, J. V. and Coleman, E. W. "Design and Use of Ten-Kilocurie Source of Gamma Radiation." Chem. Engng. Progr., 49, No. 11, 1953. 41. Nehemias, J. V., Brownell, L. E., Meinke, W. W. and Coleman, E. W. "Installation and Operation of Ten-Kilocurie Cobalt-60 Gamma-Radiation Source." Amer. J. Physics, 22, No. 2, 1954. 42. Brownell, L. E. and Nehemias, J. V. "Techniques Used in Studies with High-Intensity Gamma Radiation." Sci. Mon., N.Y., 82, No. 2, 1956.

3-58 43. Artandi, C., Stonehill, A. A. "Polyvinyl Chloride -- New High-Level Dosimeter." Nucleonics, 16, No. 5, (1958) 118. 44. Fowler, J. F. and Day, M. J. "High Dose Measurements by Optical Absorption. " Nucleonics, 13, (1955) 52. 45. Artandi, C. "Plastics Dosimetry." Nucleonics, 17, (1959) 62. 46. Alexander, P., et al. "The Degradation of Solid Polymethylmethacrylate by Ionising Radiation," Proc. Royal Soc., London, 223, (1954) 392. 47. Feng, P. Y. "Polymer Degradation -- Wide-Range Dosimeter." Nucleonics, 16, No. 10, (1958) 114. 48. Kreidl, N. J. and Blair, G. E. "Glass Dosimetry." Nucleonics, 17, (1959) 58. 49. Schulman, J. H., Shuriliff, W., Ginther, R. J. and Attix, F. H. "Radiophotoluminescence Dosimetry System of the U.S. Navy." Nucleonics, 11, No. 10, (1953) 52. 50. Anonymous, Phosphate Glass Dosimetry. Civil Defense Tech. Bull., TB-11-15, U.S. Govt. Print. Office, Wash., D.C., July, 1954. 51. Schulman, J. H., Ginther, R. J., Klick, C. C., Alger, R. S. and Levy, R. A. "Dosimetry of X-Rays and Gamma-Rays by Radiophotoluminescence." J. Appl. Phys., 22, (1951) 1479. 52. Schulman, J. H. Measurement of High Doses of Co-60 Gamma Rays by Absorption Changes in Phosphate Glass. NRL Memorandum Report 266, Naval Research Laboratories, Wash., D. C., Feb., 1954. 53. Schulman, J. H. "Measuring High Doses by Absorption Changes in Glass." Nucleonics, 13, No. 2, (1955) 30. 54. Schulman, J. H., Klick, C. C., Etzel, H. W. and Ginther, R. J. Dosimetry of Ionizing Radiations. Report of NRL Progress, Naval Research Laboratory, Wash., D.C., Oct., 1954. 55. Schulman, J. H. and Etzel, H. W. "Small-Volume Dosimeter for X-Rays and Gamma-Rays. " Science, 118, (Aug. 14, 1953) 184. 56. Kreidl, N. J. and Blair, G. E. "A System of Megaroentgen Glass Dosimetry." Nucleonics, 14, No. 1, (1956) 56.

3-59 57. Kreidl, N. J. and Blair, G. E. "Recent Developments in Glass Dosimetry." Nucleonics, 14, No. 3, (1956) 82. 58. Weiss, J. "Chemical Dosimetry Using Ferrous and Ceric Sulfate." Nucleonics, 10, No. 7, (1952) 28. 59. Weiss, J., Allen, A. 0. and Schwaxrz, H. A. Use of the Fricke FerrousSulfate Dosimeter for Gamma-Ray Doses in the Range 4 to 40 kr. Paper 155, Int. Conf. on Peaceful Uses of Atomic Energy, Geneva, 1955; Proc. ibid, United Nations, N. Y., 14, (1956) 179. 60. Fricke, H. and Morse, S. "Action of X-Rays on Ferrous Sulfate. Amer. J. Roentgenol., 18, (1927) 426. 61. Fricke, H. "The Oxidation of Ferrous-Sulfate in Aqueous Solutions by X-Rays of Different Wave-Lengths." Phys. Rev., 31, (1928)1117. 62. Fricke, H. and Morse, S. "The Action of X-Rays on FeSO4 Solutions. " Phil. Mag., Series VII, 7, (1929) 129. 63. Fricke, H. and Hart, E. J. "The Oxidation of Fe++ to Fe+++ by the Irradiation with X-Rays of Solutions of Ferrous Sulfate in Sulphuric Acid. " J. Chem. Phys., 3, (1935) 60. 64. "American Society for Testing Materials. Tentative Method of Test for Absorbed Gamma Radiation Dose by Chemical Dosimetry, ASTM Method D1671-59T." ASTM Bull. No. 239, 30 and 52, July, 1959. 65. Hochanadel, C. "Effects of Cobalt Gamma-Radiation on Water and Aqueous Solutions." J. Phys. Chem., 56, (1952) 587. 66. Mooney, R. W. and Szasz, G. J. Discussion on Radiation Chemistry. ONRL, 48-52, Office of Naval Research, Lond., May, 1952. 67. Schuler, R. H. and Allen, A. O. "Yield of the Ferrous Sulfate Dosimeter: An Improved Cathode-Ray Determination." J. Chem. Phys., 24, (1956) 56. 68. Charlesby, A. Atomic Radiation and Polymers. Pergamon Press, London, 1960. 69. Lea, D. E. Actions of Radiation on Living Cells. The Macmillan Co., Inc., N.Y., 1947. 70. Hardwick, T. J. "The Reduction of Ceric Sulfate Solutions by Ionizing Radiation."t Canad. J. Chem., 30, (1952) 23.

3-60 71. Dewhurst, H. A. "Effect of Organic Substances on the Gamma Ray Oxidation of FeSO4." J. Chem. Phys., 19, (1951) 1329. 72. Bastian, R., Weberling, R. and Palilla, F. "Spectrophotometric Determination of Iron as Ferric Sulfate Complex." Analyt. Chem., 25, (1953) 284. 73. Dainton, F. S. and Sutton, H. C. "Hydrogen Peroxide Formation in the Oxidation of Dilute Aqueous Solutions of FeSO4 by Ionizing Radiations." Trans. Faraday Soc., 49, (1953) 1011. 74. Hochanadel, C. J. and Ghormley, J. A. "A Calorimetric Calibration of Gamma Ray Actinometers." J. Chem. Phys., 21, (1953) 880. 75. Lazo, R. M., Dewhurst, H. A. and Burton, M. "The FeSO4 Radiation Dosimeter: A Calorimetric Calibration with Gamma Rays." J. Chem. Phys., 22, (1954) 1370. 76. Farmer, F. T., Rigg, T. and Weiss, J. "The Absolute Yield of the Ferrous Sulfate Dosimeter." J. Chem. Soc., (1954) 3248. 77. Hart, E. J. and Walsh, P. D. "A Molecular Product Dosimeter for Ionizing Radiations." Rad. Res., 1, (1954) 342. 78. Hart, E. J. "Radiation Chemistry of Aqueous Ferrous Sulfate-Cupric Sulfate Solutions, Effect of Gamma-Rays." Rad. Res., 2, (1955) 33. 79. Hart, E. J., et al., Measurement Systems for High Level Dosimetry. Proceedings of the Second U.N. Conference on the Peaceful Uses of Atomic Energy, United Nations, Geneva, 21 (1958) 188. 80. Johnson, M. E., Swartz, J. C. and Hamilton, A. B. Monthly Progress Reports Nos. 1-8, to the Army Chemical Corps., Physics Research Department, Vacuum Distillation Products Industries, Eastman Kodak, Rochester, N.Y., 1952. 81. Taplin, G. V., pouglas, C. H. and Sigoloff, S. C. "The ChloroformAlcohol-Dye System." UCLA-192, U.S. Atomic Energy Commission, Wash., D.C., 1952. 82. Taplin, G. V., Douglas, C. H. and Sigoloff, S. C. Quart. Progress Report 1-2, to the Army Chemical Corps.j University of California, L. A., 1952, 1953.

83. Taplin, G. V., Douglas, C. H. and Sigoloff, S. C. Gamma and X-Ray Dosimetric Method. Qlart. Progress Report 7, to the Army Chemical Corps., University of California, L. A., 1954. 84. Taplin, G. V., Douglas, C. H. and Sigoloff, S. C. Quart. Progress Report 5, to the Army Chemical Corps., University of California, L. A., 1952. 85. Taplin, G. V. Development of Direct-Reading Chemical Dosimeters for Measurement of X, Gamma, and Fast Neutron Radiation. Paper 153, Int. Conf.- on Peaceful Uses of Atomic Energy, Geneva, 1955; Proc. ibid, United Nations, N. Y., 14, (1956) 227. 86. Hilsenrod, A. "Irradiation of Chloral Hydrate Solutions." J. Chem. Phys., 24, (1956) 917. 87. Andrews, H. L., Murphy, R. E., and LeBrun, E. J. "Gel Dosimeter for Depth-Dose Measurements." Rev. Sci. Instr., 28, (1957) 329. 88. Dondes, S. A High Level Dosimeter for the Detection of Beta and Gamma Radiation and Thermal Neutrons. Paper 151, Int. Conf. on Peaceful Uses of Atomic Energy, Geneva, 1955; Proc. ibid, United Nations, N.Y., 14, (1956) 176. 89. Harteck, P. and Dondes, S. "Nitrous Oxide Dosimeter for High Levels of Betas, Gammas, and Thermal Neutrons." Nucleonics, 14, No. 3, (1956) 66. 90. Harmer, D. E. "Chemical Dosimetry." Nucleonics, 17, (1959) 72. 91. Latuente, B., Goldblith, S. A. and Proctor, B. E. "Some Further Studies on the Application of Methylene Blue in Aqueous Solution as a Dosimeter for Intense Beams of High-Energy Radiation." Intern. J. Appl. Radiation Isotopes, 3, (1958) 119. 92. Goldblith, S. A., Proctor, B. E. and Hammenle, A. O. "Evaluation of Food Irradiation Procedures." Industr. Engng. Chem., 44, (1952) 310. 93. Shekhtaman, Ya. L., et al. Doklady Akad. Nauk., USSR, 74, (1950) 767. 94. Day, M. J. and Stein, G. "Chemical Dosimetry of Ionizing Radiations." Nucleonics, 6, No. 2, (1951) 35. 95. Armstrong, W. A. and Grant, G. A. "A Highly Sensitive Chemical Dosimeter for Ionizing Radiation." Nature, 182, (1958) 747.

3-62 96. Draganic, I. "Action des Rayonnements Ionisants sur les Solutions Aquenses d'acide Oxalique: Acide Oxalique Aqueux Utilise Comme Dosimetre Chimique Pour les Doses entre 1.6 ct 160 M rads." J. Chim. Phys., 56, (1959) 9. 97. Day, M. J. "Chemical Effects of Ionizing Radiations in Some Gels." Nature, Lond., 166, (1950) 146. 98. Proctor, B. E. and Goldblith, S. M. "Oxidation-Reduction Dyes as Radiation Indicators." Nucleonics, 7, No. 2, (1950) 83. 99. Gevantman, E. H., Chandler, R. C., and Pestaner, J. F. "Tridimensional Examination of Chemical Systems Irradiated in Gel Media." Radiation Research, 7, (1957) 318. 100. Pestaner, J. F. and Gevantman, L. H. "Depth Dosimetry by Means of a Gel-Incorporated Chemical System." Radiation Research, 9, (1958) 166. 101. Miller, N. "Quantitative Studies of Radiation Induced Reactions in Aqueous Solutions. I. Oxidation of Ferrous Sulfate by X- and GammaRadiation. " J. Chem. Phys., 18, (1950) 79. 102. Miller, N. and Wilkenson, J. "Actinometry of Ionizing Radiation." Disc. Faraday Soc., 12, (1952) 50. 103. Allen, A. 0. "The Yields of Free H and OH in the Irradiation of Water." Rad. Res., 1, (1954) 85. 104. Hart, E. J., Gordon, S. and Hutchison, D. A. "Free Radical-Initiated 016 018 - H2016 Exchange Reaction in Aqueous Solution." J. Amer. Chem. Soc., 75, (1953) 6165. 105. McDonell, W. R. and Hart, E. J. "Oxidation of Aqueous Ferrous Sulfate Solutions by Charged Particle Radiations. " J. Amer. Chem. Soc., 76, (1954) 2121. 106. Barb, W. G., Baxendale, J. H., George, P. and Bargrave, K. R. "Reactions of Ferrous and Ferric Ions with Hydrogen Peroxide." Trans. Faraday Soc., 57, (1951) 462, 591. 107. Hart, E. J. "Gamma-Ray Induced Oxidation of Aqueous Formic Acid." J. Amer. Chem. Soc., 73, (1951) 68.

108. Krenz, F. H. and Dewhurst, H. A. "The Mechanism of Oxidation of Ferrous Sulfate by Gamma-Rays in Aerated Water." J. Chem. Phys., 17, (1949) 1337. 109. Rigg, T., Stein, G. and Weiss, J. "The Action of X-Rays on Ferrous and Ferric Salts in Aqueous Solutions." Proc. Roy. Soc., A, 211, (1952) 375. 110. Dainton, F. S. and Sutton, H. C. "Hydrogen Peroxide Formation in The Oxidation of Dilute Aqueous Solutions of Ferrous Sulphate by Ionizing Radiations." Trans. Faraday Soc., 49, (1953) 1011. 111. Dewhurst, H. A. "The X- and Gamma-Ray Oxidation of Ferrous Sulphate in Aqueous Solution." Trans. Faraday Soc., 49, (1953) 1174. 112. Hart, E. J. "Radiation Chemistry of Ferrous Sulfate Solutions. " J. Amer. Chem. Soc., 73, (1951) 1891. 113. Amphlett, C. B. "Reduction of Ferric Ion in Aqueous Solution by Gamma Radiation. " Nature, Lond., 171, (1953) 690. 114. Hart, E. J. "Gamma-Ray Induced Oxidation of Aqueous Formic AcidOxygen Solutions. Effect of pH." J. Amer. Chem. Soc., 76, (1954) 4198. 115. Samuel, A. H. and Magee, J. L. "Theory of Radiation Chemistry. II. Track Effects in the Radiolysis of Water." J. Chem. Phys., 21, (1953) 1080. 116. Dewhurst, H. A., Samuel, A. H., and Magee, J. L. "A Theoretical Survey of the Radiation Chemistry of Water and Aqueous Solutions." Rad. Res., 1, (1954) 62. 117. Weiss, J. "The Role of Hydrogen Molecule Ions in Aqueous Solutions." Nature, Lond., 165, (1950) 728. 118. Taplin, G. V., Douglas, C. H. and Sanchez, B. "Colorimetric Methods for Dosimetry of 10 to 100 r." Nucleonics, 9, No. 2, (1951) 73. 119. Taplin, G. V. Applicability of Chemical Dosimetry in Civil Defense. UCLA-304, U.S. Atomic Energy Commission, Wash., D.C., Sept. 15, 1955.

3-64 120. Clark, G. L. and Bierstedt, D. E. "X-Ray Dosimetry by Radiolysis of Some Organic Solutions." Rad. Res., 2, No. 3, (1955) 199. 121. Fricke, H. Army Chemical Corps Symposium No. IV: Chemistry and Physics of Radiation Dosimetry, Part I, Army Chemical Center, Maryland, 24, Sept., 1950. 122. Gomberg, H. L., Gould, S. E., Nehemias, J. V. and Brownell, L. E. "Using Cobalt-60 and Fission Products in Pork Irradiation Experiments." Nucleonics, 12, No. 5, (1954) 38. 123. Young, D. E. "Dosimetry." Sources of Radiation for Industry, IP-175 The University of Michigan, Ann Arbor, Mich., Aug., 1956. 124. Genna, S. and Laughlin, J. S. "Absolute Calibration of a Cobalt-60 Gamma Ray Beam." Radiology, 65, (1955) 394. 125. Laughlin, J. S., Genna, S., Danzker, M. and Vacirca, S. J. Absolute Dosimetry of Cobalt-60 Gamma Rays. Paper 70, Int. Conf. on Peaceful Uses of Atomic Energy, Geneva, 1955; Proc. ibid, United Nations, N. Y., 14, (1956) 163. 126. Lazo, R. M., Dewhurst, H. A., and Burton, M. "The Ferrous Sulfate Radiation Dosimeter: A Calorimetric Calibration With Gamma Rays." J. Chem. Phys., 22, (1954) 1370. 127. Taimuty, S. I., Glass, R. A. and De La Rue, R. Calorimetric Determination of the Yield of the Ceric Sulfate Dosimeter. Paper presented at third annual meeting, American Nuclear Society, Pittsburgh, June 24-28, 1957. 128. Taimuty, S. I. Electron Beam Dosimetry and Experimental Techniques. Paper presented at Symposium on Electron Beam Radiation, General Electric Co., Milwaukee, 1957; see also Nucleonic, 15, No. 11, (1957) 182. 129. Schall, P. "A Comparison of Dosimetry Methods." Nucleonics, 17, (1959) 68. 130. Birks, J. B. "Scintillations from Napthalene-Anthracene Crystals." Proc. Phys. Soc., Lond., 63A, (1950) 1044.

3-65 131. Bowen, E. J., Mikiewicz, E., and Smith, F. W. "Resonance Transfer of Electronic Energy in Organic Crystals." Proc. Phys. Soc., Lond., 62A, (1949) 26. 132. Schulman, J. H., Etzel, H. W. and Allard, J. C. "Application of Luminescence Changes in Organic Solids to Dosimetry." J. Appl. Phys., 28, (1957) 792. 133. Attix, F. H. "High Level Dosimetry by Luminescence Degradation. Nucleonics, 17, No. 4, (1959) 142. 134. Barr, N. F. and Stark, M. B. "The Destruction of the Fluorescence of Quinine in Acid Solution by 250 kvp X-Rays." Radiation Research, 9, (1958) 89. 135. Rozman, I. M. and Zimmer, K. G. 'The Damage to Plastic Scintillators byIonizing Radiations." Atomnaya Energiya, 2, (1957) 54.; English translation, Intern. J. Appl. Radiation Isotopes, 3, (1958) 36. 136. Mayernik, J. J. and Daniels, T. "The Sterilization of Polyethylene Bags by Electron Irradiation and a Bacterial Monitor as a Measure of Sterility." J. Am. Pharm. Assoc. Sci. Ed., 48, (1959) 16. 137. Brownell, L. E., "Radiation Uses in Industry and Science," U.S. Government Print Office, Washington, D. C., June 1961.

CHAPTER 4 GAMMA SHIELDING The operation of a large number of nuclear reactors, the construction of particle accelerators in the Bev range, and the use of megacurie quantities of gamma radiation in facilities using spent reactor fuel elements, cobalt and cesium sources, have made the shielding of these facilities an important branch of study. The interaction of nuclear radiations with matter has been the subject of extensive theoretical and experimental investigation, and neutron- and gamma-radiation attenuation processes have been studied in detail for the purposes of shielding. The study of shielding for a given facility involves, basically, analysis of the following factors: 1. The type of radiation source. 2. The nature of nuclear radiations involved. 3. Source strength and energy spectrum of the nuclear radiations. 4. Type of shields to be used, considering cost, weight, and type of facility. 5. The basic attenuation and spatial-distribution processes involved. 6. Evaluation of the nuclear constants of the materials involved for different energy groups. 7. Source geometry and the radiation field surrounding the source and the shield. 8. Over-all minimum shield thickness required so as not to exceed the maximum permissible dose permitted by AEC regulations. This chapter considers the calculations involved in the shielding of nuclear radiation facilities and deals primarily with gamma-radiation shielding and the concept of the "build-up factor." As an illustration, the analysis of the heterogeneous gamma-radiation spectrum from the MTR fuel element from the point of view of shielding is given. The report contains graphs for determining the standard integrals involved in calculating the radiation flux for standard geometries and also an extensive bibliography. 4.1 ATTENUATION OF GAMMA-RADIATION FROM POINT SOURCES There are no less than twelve different processes by which gamma rays can interact with matter. However, three of them, photoelectric absorption, Compton scattering, and pair production are the most important. See Chapter 2 of "Radiation Uses in Industry and Science" for greater discussion. l) The photoelectric effect is predominant for low photon energy and materials with high atomic numbers (E < 0.15 Mev for Z = 29; E 0.5 Mev for Z = 82). At energies greater than 1.02 Mev, pair production predominates (E > 80, 15, or 5 Mev for Z = 1, 13, or 82, respectively). For intermediate energies, Compton scat4.1

tering is predominant. The cross sections per atom for the photoelectric and Compton effectsand pair-production process vary approximately as Z5, Z, and Z(Z+1), respectively. Materials of high atomic numbers are more suitable as absorbers in the low-energy region, while they are not quite as efficient in the high-energy region. Lead, zinc bromide, concrete, and water are the materials commonly used for shielding, the actual choice depending on space, weight, and cost considerations. In fixed-source facilities, special types of concrete and water solutions are frequently used. The stream of photons emitted from the source in all directions decreases in intensity inversely as the square of the distance for a point source. This relationship can be expressed simply as Loaffi~~~~ rG!(4.1) Here 0 is flux, r is distance, and 00 is a measure of the strength of the source (see Fig. 4.1). Equation 4.1 is referred to as the "inverse-square law." Thus, if space is available, the problem of- gamma shielding can be reduced by increasing the distance of closest approach permitted. The above equation applies to trasmission through a vacuum or a thin gas. The penetration of photons in material media of greater densities is now considered. Assume a point monoenergetic source of gamma rays in vacuum is shielded from a point detector located at a distance R by a rod of some material, of length X, and oriented as shown in Fig. 4.2. The other dimensions will be assumed infinitesimally small relative to X, R, and (R-X). In such a configuration every collision which a photon from the source experiences in the rod will then serve to remove it from the beam which ultimately reaches the detector. Such an arrangement of source, shield, and detector is termed a "good geometry." In such a case, the rate of reduction of flux is proportional to the flux level, i.e., d/dx = - (4.2) or f= O Xe x (4.3) Here 0o is the flux level entering the rod, I is the flux level existing at a point in the rod at a distance x from the "inside" end, and AL is a constant (the "attenuation coefficient") the value of which depends on the photon energy and the particular material constituting the rod. The dimension of t is cm-1l so that. times the thickness of the absorber in cm is a dimensionless constant which termed the thickness in relaxation lengths." Equations 4.2 and 4.3 are valid only if the gamma rays are monoenergetic and the beam is collimated 4.2

Fig. 4.1. Flux reduction with distance. 4.3

ig. x o-gc Fig. 4.2. Good-geometry configuration. If the source emits isotropically, then for a narrow rod absorber and a point detector, a combination of Equations 4.1 and 4.3 may be used to express 0 as 4 e-X r2 Now if the diameter of the rod is not small, the arrangement is termed "bad geometry." Here photons can be scattered into the detector as well as away from it (see Fig. 4.3). Equation 4.4 does not properly describe gamma attentuation in "bad geometries," but must be corrected for the effect of the scattered photons. OURC - _ __-.- DETECTOR 1I- x 1 —x Fig. 4.3~. Bad-geometry configuration. 4 4

The total attenuation coefficiejt (a function both of energy and absorber material) is (Eo) = (E) + a(Eo) + K,(Eo), (4.5) where T, a, and K denote the photoelectric, Compton, and pair-production coefficients, respectively. Of these three interaction processes, the photoelectric effect and pair-production may be treated as purely absorptive. The third, Compton scattering, is the only scattering process of the three and gives rise to a spectrum of photons degraded in energy. 4.2 NARROW BEAM ATTENUATION Figure 4.4 shows a plot of the linear attenuation coefficient for "narrow beam" (collimated) gamma radiation in lead.(2) The three components T, a, and K corresponding to the three processes of interaction are also shown in Fig. 4.4 and the dependence of each upon the energy of the radiation is apparent. As previously pointed out, attenuation by photoelectric effect, T, is most important in the low energy region, particularly for the heavy elements. Figure 4.4 shows that below 0.5 Mev the primary contribution to p. is from T. However the curve of T versus Mev drops very rapidly in this range with the result that above 1.0 Mev there is very little attenuation of gamma radiation by photoelectric effect. At medium energies of from 0.5 to 5 Mev Compton scattering a is the most important attenuation process. For the heavy elements Compton scattering is less important below 0.5 Mev than the photoelectric effect but for the lighter elements the relative importance of scattering continues to lower energies. In using Equation 4.5 it is not necessary to express the absorber thickness in units of length. Sometimes it is more convenient to use mass, especially in considering a non-homogeneous absorber. In such a case x may be expressed in gins per sq cm or in lb per sq ft. The attenuation coefficient is then termed the mass attenuation coefficient ~m and has the units of sq cm per gm or sq ft per lb respectively. The simple relationship ptm = p/p where p is the density of the absorber is conveneint in relating these coefficients. Figures 4.5, 4.6, and 4.7 give mass attenuation coefficients for the photoelectric range, the Compton scatter range and the pair-production range respectively for some of the common elements as a function of energy.(354) Table 4.1 gives values for some common elements, air and water as a function of photon energy.(3) 4.3 HALF-VALUE AND TENTH-VALUE THICKNESSES Instead of using attenuation coefficients it is often more convenient to express the attenuation in terms of half-value layers or tenth-value layers.

U I.Z 0.6 z w 0.5 UL. w 0 z 0 cl,~~~~~~~~~~~i i w 0.2 Iw z 0.. 0~... Iz o [ 0 I 2 3 4 5 6 7 8 9 10 II 12 ENERGY IN MEV Fig. 4.4. Linear attenuation coefficient of gamma rays in lead (2)

1.0 0.9 0.8 - 0.7 0.6 -L 0.5 CARBON ARWTER LUMINUM COPPER LE M IIRd E o 03 0.2 E.I.0 1.02.03.04.06.08 0.1 2.3.4.6.8 1.0 ENERGY IN MEV Fig. 4.5. Total mass attenuation coefficients, PLm, for x- (ayd gamma radiation in the range of photoelectric effect. II, 7

.10 _ _ Pb 'O.08 c.07 N~~~~~~~~~~~~~ 06 E 0 06......._'.... 4.0.03 o0.3 0.4 0.5 0.6 0.8 1.0 1.5 2.0 ENERGY IN MEV Fig. 4.6 Total mass attenuation coefficients, )'m' for x- and. gamnmi radiation in the range of Comnpton scatter. (3) II~~~~~~~~~~~~~~~~~~~~~~~~~~~I.03~~~~~~~~~~~~~~~~~~~'" 0.3 0.4 0.5 0.6 0.8 1.0 1.5 2.0 3.0 ENERGY IN MEV Fig. 4.6 Total mass attenuation coefficients, ~m, for x- and gamma radiation in the range of Compton scatter. (3)

.10 loJ E____~~~~~~~~~~~~~~~~, _,, l l llP b.08 I, a- Ai.07 Fe,,,,,,,,'A 1.0 2.0 3 4 5678910 20 3040 6080100 E.03.0 1 Fig. 4.7 Total mass attenuation coefficients.,, for x- and gamma

TABLE 4. 1 GAMMA RAY MASS ATTENUATION COEFFICIENTS, Vm, WITHOUT COHERENT SCATTERING, IN cm2/g Material E, Mev H Be C O Na Al Si Fe Air H20 0.01 0.385 0. 533 2.13 5.69 15.6 26.3 34.1 178 4.88 5.09 0.015 0.376 0.261 0.701 1.68 4.61 7.84 10.3 58.2 1.48 1.53 0.02 0.369 0.200 0.382 0. 766 1.95 3.33 4.35 25.8 0. 695 0.720 0.03 0.357 0.168 0.229 0.334 0.646 1.04 1.35 8.03 0.317 0.336 0.04 0.345 0.158 0.193 0.232 0.351 0.507 0.633 3.48 0.226 0.245 0.05 0.335 0.151 0.178 0.196 0.249 0.325 0.389 1.83 0.194 0. 212 0.06 0.326 0.146 0.169 0.179 0.206 0.249 0.288 1. 13 0. i78 0.196 0.08 0.309 0.138 0.157 0.161 0.168 0.186 0.204 0.555 0.161 0.178 0.10 0.294 0.132 0.149 0.151 0.151 0.161 0.172 0.344 0.151 0.167 0.15 0.265 0.118 0.134 0.134 0.130 0.133 0.139 0.183 0. 134 0.149 o 0.20 0.243 0.109 0.122 0.123 0.118 0.120 0.125 0.138 0.122 0.136 0.30 0.211 0.0944 0.106 0.106 0.102 0.103 0.107 0.106 0.106 0.118 0.40 0.189 0.0846 0.0953 0.0954 0.0912 0.0922 0.0956 0.0919 0.0952 0. 106 0.50 0.173 0.0772 O. 0870 0.0871 0.0833 0.0840 0.0869 0.0828 0.0869 0. 0967 0. 60 0. 160 0.0714 0.0805 0.0805 0.0770 0.0777 0.0804 O. 0761 0.0804 0.0894 0.80 0.140 0.0628 0.0707 0.0708. 0.0677 0.0682 0.0706 0.0664 0.0706 0.0786 1.0 0.126 0.0564 0.0635 0.0636 0.0608 0.0613 0.0635 0.0595 0.0635 0.0706 1.5 0. 103 0.0459 0.0518 0.0518 0.0496 0.0500 0. 0517 0.0484 0.0516 0.0576 2.0 0.0876 0.0393 0.0443 0.0445 0.0427 0.0432 0.0447 0.0424 0.0443 0.0493 3.0 0.0691 0.0313 0.0356 0.0359 0.0348 0.0353 0.0367 0.0360 0.0357 0.0396 4.0 0.0579 0.0265 0.0304 0.0309 0.0303 0.0310 0.0323 0.0330 0.0307 0.0339 5.0 0.0502 0.0233 0.0270 0.0276 0.0274 0.0282 0.0296 0.0313 0.0274 0.0301 6.0 0.0446 0.0211 0. 0245 0. 0254 0.0254 0.0264 0.0277 0.0304 0.0250 0.0275 8.0 0.0371 0.0180 0.0213 0.0224 0.0229 0.0241 0.0255 0.0295 0.0220 0.0240 10.0 0.0321 0.0161 0.0194 0.0206 0.0215 0.0229 0.0243 0.0295 0.0202 0.0219

A half-value layer is the thickness.of an absorber that will absorb 1/2 of a beam and similarly a tenth-value layer is the thickness of absorber that will transmit 1/10 of a beam. The following relationships exist: tl = half-value thickness = 0.693/h (4.6) 2 0 =, where n = number of half-value thicknesses (4.7) 2n Figure 4.8 shows some 1/10 value thicknesses (narrow beam)(5) for some common materials as a function of energy. 4.4 MIXED ENERGIES In the case of mixture of gamma emitting radioisotopes such as the fission products, gamma radiation of a mixture of different energies is obtained. However, the attenuation coefficients and half- and tenth-value thicknesses given are constant only for a given absorber and energy. When a mixture of energies is involved the decrease in the flux of a collimated beam is given by:(6) = 01 e-I lx + 2 e-i2x +.. e Anx (4.8) where: 01, 02,...-n = the flux at the surface of the absorber from each of n-radiation energies f1t, 12Y -n = attenuation coefficients for the corresponding radiation energies. If a large number of different energies are involved or if the energy is continuous in distribution approximate methods or laboratory measurements are required,. 4.5 ATTENUATION "BUILD-UP" FACTORS FOR POINT SOURCES In the ideal case of absorption of a perfectly collimated beam the degradation of energy, scattering, and the production of secondary radiation can be eliminated from consideration. But in most real applications the "build-up" of secondary radiation is an important consideration. The radiation intensity at any point in an absorber consists of the primary gamma photons, the photons resulting from Compton scatter, and the secondary radiation of (photoelectrons, and electrons from Compton scatter and pair production) plus the x radiation produced by the absorption of these electrons. The secondary electrons are absorbed much more readily than the primary gamma radiation. 411

0100. 80 0 60 0 40 O'-CC)~~~~~~~~~~~O >. 20 0 60.08 ~~0 ~~ ~~~',14 w.08 o.06.04.02.01.1.2.4.6.8 1 2 4 6 8 10 ENERGY (MEV.) Fig. 4.8. Narrow beam tenth value thicknesses of various materials for gamma radiation. (5) 4. 12

For primary gamma radiation of.a given energy passing through a given absorber an equilibrium level is reached in the production of secondary radiation which is characteristic of the energy and the absorber. If the primary radiation passes from a poor absorber such as air to a denser medium the increase in absorption of the primary radiation will result in a corresponding increase in the formation of secondary radiation. This produces a "build-up" of secondary electrons and x radiation from the absorption of these electrons. Figure 4.9 is a diagram (not to scale) showing such a "build-up" of secondaries as a primary photon passes through a denser medium.(6) Intensity ' (not to scale) Primary + Secondary / AirJ Lead ir Fig. 4.9. Diagram of "build-up" in lead. (6) Tables 4.2, 4.3, 4.4, and 4.5 give the experimentally determined dose build-up factors B from a point source for H20, Al, Fe and Pb respectively.(7) In using these factors the calculated flux intensity 0 based upon absorption of a collimated beam is multiplied by factor B or: F = B,0 e dx (4.9) 4.13

TABLE 4.2 DOSE BUILD-UP FACTOR B IN WATER FOR AN ISOTROPIC POINT SOURCE(7) (Relaxation lengths, px) EO(Mev) j 1 2 4 7 10 15 20.256 1 3.09 7.14 23.0 72.9 166 456 932.5 2.52 5.14 14.3 38.8 77.6 178 334 1.0 2.13 3.71 7.68 16.2 27.1 50.4 82.2 2.0 1.83 2.77 4.88 8.46 12.4 19.5 27.7 3.0 1.69 2.42 3.91 6.23 8.63 12.8 17.0 4.0 1.58 2.17 3.34 5.13 6.94 9.97 12.9 6.0 1.46 1.91 2.76 3.99 5.18 7.09 8.85 8.o0 1.38 1.74 2.40 3.34 4.25 5.66 6.95 10.0 1.33 1.63 2.19 2.97 3.72 4.90 5.98 TABLE 4.3 DOSE BUILD-UP FACTOR B IN ALUMINIUM FOR AN ISOTROPIC POINT SOURCE(7) (Relaxation lengths, tax) E0o(v) 1 1 2 4 7 10 15 20.5 92.37 4.24 9.47 21.5 38.9 80.8 141 1.0 2.02 3.41 6.57 13.1 21.2 37.9 58.5 2.0 1.75 2.61 4.62 8.o5 11.9 18.7 26.3 3.0 1.64 2.32 3.78 6.14 8.65 13.0 17.7 4.o 1.53 2.08 3.22 5.01 6.88 10.1 13.4 6.0 1.42 1.85 2.70 4.06 5.49 7.97 10.4 8.0 1.34 1.68 2.37 3.45 4.58 6.56 8.52 10.0 1.28 1.55 2.12 3.01 3.96 5.63 7.32 TABLE 4.4 DOSE BUILD-UP FACTOR B IN IRON FOR AN ISOTROPIC POINT SOURCE(7) (Relaxation lengths, px) Eo (v) 1 1 2 4 7 10 15 20.5 1.98 3.09 5.98 11.7 19.2 35.4 55,6 1.0 1.87 2.89 5.39 10.2 16.2 28.3 42.7 2.0 1.76 2.43 4.13 7.25 10.9 17.6 25.1 3.0 1.55 2.15 3.51 5.85 8.51 13.5 19.1 4.0 1.45 1.94 3.03 4.91 7.11 11.2 16.0 6.0 1.34 1.72 2.58 4.14 6.02 9.89 14.7 8.o 1.27 1.56 2.23 3.49 5.07 8.50 13.0 10.0 1.20 1.42 1.9 2.99 4.35 7+ 54 12.4 4.14

TABLE 4.5 DOSE BUILD-UP FACTOR B IN LEAD FOR AN ISOTROPIC POINT SOURCE(7) (Relaxation lengths, px) Eo (v) J1 2 4 7 10 15 20.5 1.24 1.42 1.69 2.00 2.27 2.65 2.75 1.0 1.37 1.69 2.26 3.02 3.74 4.81 5-.86 2.0 1.59 1.76 2.51 3.66 4.84 6.87 9. 00 3.0 1.34 1.68 2.43 3.75 5.30 8.44 12.3 4.0 1.27 1.56 2.25 3.61 5.44 9.80 16.3 5.1097 j 1.21 1.46 2.08 3.44 5.55 11.7 23.6 6.0 1.18 1.40 1.97 3.34 5.69 13.8 32.7 8.0 1.14 1.30 1.74 2.89 5.07 14.1 44.6 10.0 1.11 1.23 1.58 2.52 4.34 12.5 39.2 4.6 "BROAD-BEAM" COEFFICIENTS Another method of calculation is to use coefficients experimentally determined for a "broad" beam rather than a narrow (collimated) beam. Figures 4i0, 4.11, and 4.12 give transmission in concrete, Fe and Pb respectively of primary radiation from Ra, Co-60, and Cs-137 with corrections for secondary radiation. These Figures are based on "broad" beam coefficients. (8) A necessary quantity for shielding calculations is the specific radiation flux 0o given as a function of the energy of the emitted gamnma radiation. This quantity is plotted in Fig. 4.13 as roentgens per sq cm at one centimeter distance per photon emitted per disintegration per millicurie of radioisotope.(9) 4.7 ENERGY ABSORPTION COEFFICIENTS, 'La The previous sections discuss the attenuation of primary radiation flux from point sources and the build-up of secondary radiation. The decrease in the number of primary photons may be determined by use of narrow beam attenuation coefficients and corrections may be made for the attenuation of secondarry radiation by use of broad.a beam coefficients or build.-up factors. However these proced~ures d~o not give the amount of energy absorption in the medium under consideration. Customary units of dose (roentgen, rep and rad) are units of energy absorp4.15

FRACTION of TRANSMISSION o o 0 0 b o o o b ' 0 0 0 0 c O 0 00 b v00 09 CO~ C) 0; I i I o Z 0 n 0D C/) O 0 ~g H 0 _ 0 0 PI — 0 "

LT 1~ (8) UOJT UTF sxeJ 'emrmeS L[-1nTSGOa puBe 09 —Teaqoo 'ImfpaPa: jo uoTss'uIsue aq,Tleaq-pteota:T 1 ' TT't 'JT (WO) NO8l }0 SS3N>OIH. OL 09 0 Ot 0t 0 OZ OI O. 10000' I IL\\ \ II 1000 ___ 0903 ~'~ '~.. _10 "L I000' \\ z -' R10' u _.-01 "~~~~~~~~~~~~. 1

ob~~ ~FRACTION of TRANSMISSION o b 0 0 b 0 o 0 b~ -CQ MON ~....~~~~~~~~~~~~~ 0 000"W ~D ps 0 *m m (t~~~~,t 0 wmwm mem t m0 H 0 Oq o 0 In P) (D Ft A Pi 0 (E) m o O.p. PI a~~~~~~~~~~~~ I

z w -- aw 10 U. _ I -= T - WI.001.01 0.1 I 10 Fig. 4.13. Specific radiation flux, ) as a function of energy in Mev of gamma radiation.(9)

tion (as defined in Chapter 1) rather than units of photon flux. Table 4.6 gives values of energy absorption coefficients, pa, for some common elements, air and water as a function of the energy of the primary radiation. A comparison of corresponding values for energy absorption and flux attenuation can be made from Tables 4.1 and 4.6. For radiation in the range from 100 kev to several Mev the energy absorption is approximately proportional to Z/A for the absorber. For the case of air this ratio is about 0.50 whereas for water and tissue this ratio is about 0.55. Thus if unit masses of air and water are exposed to the same radiation flux for the same length of time the water will absorb 0.55/ 0.50 times as much energy as the air. That is if the air is given a dose of one roentgen (83.8 ergs per gram of air) the water will receive 83.8 (0.55/ 0.50) or about 93 ergs per gram of water. This unit of dose to water or tissue has been called the rep. Since materials other than water and tissue have different energy absorption coefficients the use of the rep as a unit has caused considerable confusion and has been discontinued in favor of the more general unit, the rad which is equal to 100 ergs per gram of absorber. The rad at present is the preferred unit of radiation dosage and is equal to same amount of energy absorbed per unit mass (100 ergs/gm) for all materials. If the radiation flux (photons per sec), the exposure time, and the energy of the radiation are known the dose in rads may be calculated by use of Table 4.6. 4.8 EXAMPLE PROBLEM1, POINT SOURCE An example, Problem 1 will demonstrate the use of the experimentally determined total mass attenuation coefficient, build-up factor, per cent transmission and tenth value thickness, in determining the change in flux due to an isotropic point source when a material shield is interposed. Example 1: S p S = 1000 CURIE Co-60 ____- _____/ f POINT SOURCE P = DETECTOR 90 Cm 15 Cm SHIELD =15 Cm STEEL STEEL Fig. 4.14. Configuration for example Problem 1. 4. 20

TABLE 4L.6 ENERGY ABSORPTION COEFFICIENT, 1ap FOR GAMM4A RAys,(1) IN cm/r; (2) IN THOMPSON UNITS/ELECTRON(10) H20 Air Al Fe Sn W Pb U E, Mev (1) (2) (1) (2) (1) (2) (1) (2) (1) (2) (1) (2) (1) (2) (1) (2) 0.01 4.89 22. 0 4. 70 23. 5 26.1 135 178 953 0.015 1.33 5.96 1.30 6.48 7.67 39.7 58. 0 311 0.020 0. 520 2. 34 0. 516 2. 58 3.15 16.3 25. 6 137 0. 030 0.15 0. 66 0.15 0. 74 0. 877 4. 54 7.87 42. 2 41.3 245 0. 040 0.064 0. 29 0. 064 0. 32 0. 351 1.82 3. 34 17. 9 18. 7 111 0.050 0.0396 0.178 0. 0388 0.194 0.17 0. 90 1. 68 9.03 10.1 59. 7 0.060 '0.0282 0.127 0. 0270 0.135 0.10 0. 54 0. 990 5. 31 6.14 36.4 0.080 0.0256 0.115 0.0238 0.119 0.0535 0.277 0. 427 2.29 2. 75 16. 3 7. 38 45.8 0.10 0.0251 0.113 0.0234 0.117 0.0378 0.196 0. 224 1.20 1.47 8. 73 4.11 25. 5 5.20 32. 8 0. 961 6.21 0.15 0. 0278 0.125 0. 0250 0.125 0. 0278 0.144 0. 0813 0.436 0.472 2.80 1.35 8. 39 1. 76 11.1 2. 34 0.20 0.0298 0.134 0.0264 0.132 0.0276 0. 143 0. 0489 0.262 0.223, 1. 32 0. 630 3. 91 0.819 5.17 1.10 708 0. 30 0.0318 0.143 0. 0284 0.142 0. 0280 0.145 0. 0336 0.180 0.0876 0. 519 0.229 1.42 0. 295 1.86 0. 392 2.53 0.40 0.0331 0.149 0. 0296 0.148 0.0290 0.150 0. 0308 0.165 0.0536 0. 318 0.12 0.75 0.16 0. 98 '0.207 1.34 0. 50 0. 0329 0.148 0. 0296 0.148 0.0286 0.148 0.0293 0.157 0.0401 0.238 0. 0787 0.488 0.0995 0.828 0.13 0.85 Fj 0.60 0.0329 0.148 0. 0296 0.148 0.0286 0.148 0. 0287 0.154 0.0346 0. 205 0. 0601 0.373 0.0737 0.465 0.097 0.625 0.80 0.0322 0.145 0. 0288 0.144 0.0278 0.144 0. 0274 0.147 0.0295 0.175 0. 0426 0.264 Q.0504 0.318 0.0628 0.406 1.0. 0.0311 0.140 0. 0280 0.140 0. 0268 0.139 0. 0263 0.141 0.0268 0.159 0.0353 0.219 0.0403 0.254 0.0481 0.311 1.5 0. 0287 0.129 0. 0256 0. 128 0.0249 0. 129 0. 0242 0.130 0.0240 0.142 0.0282 0.175 0.0306 0.193 0.0347 0.224 2. 0 0. 0265 0.119 0. 0238 0.119 0.0234 0.121 0. 0231 0.124 0.0233 0.138 0. 0271 0.168 0.0293 0.185 0.0325 0.210 3. 0 0. 0233 0.105 0. 0212 0.106 0. 0212 0.110 0. 0224 0.120 0.0245 0.145 0. 0287 0.178 0.0306 0.193 0.0331 0.214 4.0 0.021 0.095 0. 019 0. 097 0. 0201 0.104 0. 0224 0.120 0.0258 0.153 0.0310 0.192 0.0328 0.207 0.0351 0.227 5. 0 0.020 0. 089 0. 018 0. 091 0. 0193 0.100 0. 0228 0.122 0.0277 0.164 0.0335 0.208 0.0353 0.223 0.0375 0.242 6. 0 0. 019 0. 085 0. 017 0.086 0.0189 0. 098 0. 0231 0.124 0.0292 0. 173 0. 0355 0.220 0. 0372 0.235 0. 0395 0.255 8. 0 0. 017 0. 078 0. 018 0.080 0.0183 0. 095 0. 0239 0.128 0.0317 0.188 0.0390 0.242 0.0412 0.260 0.0432 0.279 10.0 0. 0165 0.0743 0.015 0. 077 0.0183 0. 095 0. 0250 0.134 0.0342 0. 203 0. 0426 0.264 0.0452 0.285 0.0474 0.306

A. What is flux at point P if no shield is used? 1. Base calculations on 0o (spec. rad. flux) (see Fig. 4.13). 2. Base calculations on photons emitted. And ka = 0.027 cm2/gm (see Table 4.6) give answers in r/hr. B. What is flux at point P with shield? 1. Base calculatior on % transmission. 2. Base calculations on 1/10 value thickness. 3. Base calculations on narrow beam coef. pm = 0.054 cm2/gm (see Table 4.1). 4. Base calculations on build-up factor B + an, P steel = 7.78gm/cc. Solution: A. 1. From Fig. 4.13 using avg Mev Co-60 - 1.17+1.38 = 1.25 Mev 0o = 7 r/hr per mc at 1 cm (per photon) = 7 r/hr (2 photons) 103 curies photon05 cm photon (10-3 curiesi ) 1 cm 1 cm 1.27 x 103 r/hr (without shield). 2. 103 curies (35.7 x 1010 dis/sec - curie)(2 phots/dis) 4t (105 cm)2 5= 34 x 108 photons/sec - cm2 5.34(10) 8Z/sec (1.25 Mev) (1.6 x 10 ergs) (3600 sec) cm2 y Mev hr 4.22

= 3.83 x 106 ergs/hr - cm2 Md_ = =- m = 0.027 cm2 (3.83 x 106 ergs/hr - cm2 dx gm(air) 83.8 ergs/gm (air) r = 1.24 x 103 r/hr B. 1. From Fig. 4.11 fraction transmission = 0.012: = 0.012 (1.24 x 103 r/hr) =14.9 r/hri (From broad beam calculations.) 2. From Fig. 4.8: tro = 2 in, n = 15cm 2.95 t1o 2.54 cm/in (2 in/tlo) = 0/tl = 1.24 x 10o r/hr 1.24 x 10s r/hr 102.95 890 13.59 r/hr (Obviously narrow beam calculations.) 3. 0 = 0O e t = 0.054 cm2/gm (7-78 gm/cc) = 0.420 cm-' tx = 0.42 cm-1 (15 cm) = 6.3 0 = (1.24 x 103 r/hr)e-6.3 = 1.24 x 103 (0.00183) 2.27 r/ hr 4.23

4. From Table 4.3 For AIx = 7 at 1 Mev B = 10.2 2 Mev B = 7.25 2.95 x 0.25 = 0.739 B1.25 Mev = 10.2 - 0.74 = 9.46 For Ax = 4 at 1 Mev B = 5-39 2 Mev B = 4.13 1.26 x 0.25 = 0.315 B1.25 Mev = 5.39 - 0-32 = 5.07 For Atx = 6.3 at Ax = 7, B1.25 = 9.46 vX = 4, B1.25 = 5-07 2 -4.39 x = 3-37 ix = 6.3, B1.25 = 5.07 + 3-37 = 8.44 q = B o0 e X = 8.44 (1.24 x 103 r/hr)e6e'3 8.44 (2.27) r/hr 19.2 r/hr 4.9 CALCULATION PROCEDURES FOR VARIOUS GEOMETRIES AND MULTIPLE SHIELDS Various methods of calculation have been devised to solve the problem of penetration of gamma rays in "bad geometry" situations. One such method employs the "build-up factor," B, which, used as a multiplier in Equation 4.4, corrects for the scattering effect listed above: = B o e2X (4.10) 4.24

For an isotropic point source of strength so, the inverse square law may be expressed as 0 = so/4Tr2, since all the flux is emitted isotopically in a solid angle of 4mT radians. Then Equation 4.6 can be expressed as B So -~Lx 0- e2L (4.11) 4tr2 The build-up factor introduced in Equation 4.10 has been defined as the ratio of total intensity from scattered plus unscattered photons to the intensity from unscattered photons only. Several build-up factors based on number of photons, energy transmission, dose rate, and energy-absorption factors have been defined. Assume that one photon per second is being emitted from an isotropic point source. Let the scattered flux density measured by an isotropic detector insensitive to the unscattered photons be equal to 6,s At a distance r from the source in an absorber of absorption coefficient L corresponding to source energy Eo. Various build-up factors can then be defined as follows. Types of Build-Up Factors 1i. Number Build-Up Factor, BN BN(r,Eo) = 1 + scattered photon flux unscattered photon flux BN(r,Eo) 1 + 0s ' 4cr2 er'. (4.12) 2. Energy Build-Up Factor, BE(r,Eo) BE(r,E0) = 1 + scattered energy flux unscattered energy flux The energy flux, I, is a function of r and E, and I (r,E) = Eos(r,E) (4.13a) BE = 1 + 1 4rr2 e r I(r,E)dE (4.13b) Eo where Eo is source photon energy and E is scattered photon energy. Similarly, dose rate and energy-absorption factors are defined. 4.25

3. Dose Build-Up Factor, Br (in air) Dose intensity = (photon flux) x (energy) x (absorption coefficient). 1 ~r air Br = 1 + air J4jrr2 e r Aa (E) I(r,E)dE, (4.14) EoIta (B0) where taair(Eo) is the absorption coefficient of dry air at energy Eo and ahere ~Ia air Iaair(E) is the absorption coefficient of dry air at energy E. Since pta is practically constant with energy, energy and dose build-up factors are practically the same, i.e., BEa-Br. 4. Energy-Absorption Build-Up Factor, BA(in absorbing medium) BA = 1 + 1 4r2 e e r Pka(E) I(r,E)dE (4.15) Eon ca(Eo) is the absorption coefficient for material at energy Eo. This build-up factor is important for heat-absorption calculations. 4.10 ONE-MAITERIAL SHIELD From the safety point of view, the dose build-up factor Br is most important. Extensive numerical calculations using isotropic point sources and monodirectional plane sources in infinite media have been carried out by Goldstein and Wilkins(() for lead, iron, aluminum, and water in the energy range of 0.5-10 Mev using the "Standard Moments Method" developed by Spencer and Fano.(ll) The build-up factors obtained by Goldstein and Wilkins are given in Tables 4.1 - 4.4. An empirical formula consisting of two exponentials in three parameters has been obtained by Taylor(l2) to fit the data on build-up factor for point isotropic sources in infinite media obtained by the NBS-NDA group.(7) In a simple form, the build-up factor for a point source in an infinite homogeneous medium can be expressed as the sum of two exponentials: B(pXx,Eo) = Al(Eo) exp [- j1(Eo) Ax] + A2(Eo) exp [ - a2(Eo) ~ix] (4.16) Al, An, cl, and U2 depend on energy, material of absorption, and type of build-up factor and are given in Figs. 4.15 - 4.18 for lead, iron, concrete, and water, respectively. From the definition of build-up factor, Al + A2 = 1, and p(Eo) is the narrow-beam absorption coefficient of the absorber for an 4.26

3.0 25 uL x limits:0 to 20 2.5 -I=I-A2 -Zl 2.0 - 20 x limits: Oto15 3.5 B2MO/ X)=A e U X+ A;CI6 X 1.5.35 z 3.0 o 1.0.151 L..30 t(2 bt W Q U..Q 2.5 Z o 05 W, Al I -A-.25 oo.2 _ Ui~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~' —3 0 <L2.0 n" 0 W V)~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I ~~~~~~~~~~~~~~~L..J 0.0 <~ 2.0... 0 I 0 0 oo -.20 LU (J 0 1.5 U) -0.5 m.15 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~_______________ _ ____.___ 0 1.0 ~ ~ ~ ~ ~ ~ ~ ~ ~~- -1.0.I 0 0.5 r III1-1.5.05 X 6,X BI (Eo,p. x)=Aie[P + Age I -2.0 a o 2 4 6 8 10 0 2 4 6 8 1o INITIAL GAMMA ENERGY, MEV INITIAL GAMMA ENERGY (MEV) Fig. 4.15a. Dose build-up factor in lead for Fig. 4.15b. Energy-absorption build-up factor in isotropic point source.(13) lead for isotropic point source.(3l5)

12 16 11 15 - 14 Ex B2(Eo,p~x ) =Al e-alp. + + A2e-a2,uAx 9 - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' 0.09 1..8~ 0.08 N~~~,6 7 0.07 _ DO w U IL_ _ _ _ _ __,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-," co <Sr u) - 6 0.06 < - o~ 4Uo 0.04 8 4 2 0.02~~~~~~~~~~~~~~~~~~~~~~~~~~ I"O~~~~~~~~~~~~~~~.. CL/,____ ___ - -.,01 0.05 2 07 0 4 002 344 6 70.06Ic INITIAL GAMMA ENERGY) MEV - - 0 I 23456 7 8 910 INITIAL.GAMMA ENERGY, MEV Fig. 4.16a. Dose build-up factor in iron for Fig. 4.l6b. Energy-absorption build-up fa isotropic point source.(13) iron for isotropic point source.(h13)

12 (4 (/ X+A x 0.12 B2(Eo,P. x)=Ale~,J X+A II 0.11 10 O.1 02 9 0.09 8 0.08 o a:: 7 0.07 0o ~ o LL~~~~~~~~~~~~~~-I LL w 6 '-0.06 <:1 u) 5 0.05 O U) 4 0.04 AI=I-A2 2 0.02 / 0.01 0 0 0 I 2 3 4 5 6 7 8 9 10 INITIAL GAMMA ENERGY, MEV. Fig. 14.17. Dose build-up factor in concrete (sp. gr. = 2.3) for isotropic point source.(13) 4.29

24 I I 24 22 22 20 20 B3(Eo,/ x):A;le X+ A, x i8 e2E Bo(E,/ x)Ale' +A=,eAu. X 18 16 16,Q Z 14014 14 0.14 1 (2 2 V 0 1 2 0.12 1 21.1 20~~~~~~~~( w o - ~~~~~~~~~~~~~~~~~~~~~~3I0 w l o _ 0.1 o ) l _ _ 0.08 8 0.08 0.06 6 -, 0 0 -(I A- 0.04 4~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~A= 2 ~~~~ ~ ~~~~~~~~~~0.02 2A=I-200 0 0_0 I 2: 4 5 6 7 8 9 I0 I 2 3 4 5 6 7 8 9 I0 INITIAL GAMMA ENERGY, MEV INITIAL GAMMA ENERGY, MEV Fig. 4.18a. Dose build-up factor in water for Fig. 4.18b. Energy-absorption build-up factor in isotropic point source.(13) water for isotropic point source.(13)

unscattered photon energy Eo. An advantage of using this analytical expression is the simplicity of integrating the attenuation kernel, which stays exponential, though modified by the build-up factor. 4.11 SEVERAL SLABS OF DIFFERENT MATERIALS In the case of composite shields, several different methods for using the build-up factors have been proposed. 1. For a light material, followed by a heavy material, the build-up factor of the heavy material only is used. For example, water followed by the concrete, the build-up factor of the concrete must be employed. 2. For a heavy material followed by a light material, the product of the individual build-up factors is taken. For more than two materials, this leads to conservative, high results. 3. The build-up factor may be weighted according to the number of relaxation lengths of each material. 4. The shield may be converted to an equivalent shield of known material and the build-up factor for that material may be used. This conversion may be obtained by assuming that the radiation flux, 0, on the source side of the shield are the same in both cases. OoB1 e, ~ztl = AoB2 e-e2t2 (4.17) where = the radiation flux on the source side of the shield (assumed to be plane monodirectional), 1 and 1t2 = attenuation coefficients for the materials represented by suffixes, and tl and t2 = thicknesses of the shields. Taking logarithms of both sides, ltl - k2t2 = ~n B1 B2 tz = 2 t2 + 1 (~n B1). (4.18) 4.3 41 B2 4531

Equation 4.18 can be solved for p.ltl by an iterative process. As a first approximation, assume pltl = p2t2, and use the build-up factor B1 for that value of Altl and solve for tl. The process is repeated to obtain the second approximation, using the build-up factor B1 for the calculated value of kiltl. Generally two approximations are sufficient to obtain the equivalent thickness tl in terms of the known material thickness t2. (15) 4.12 EFFECT OF GEOMETRY (FROM ROCKWELL) Equations for the gamma-radiation flux from sources of standard geometry such as point, line, disc, cylinder, and sphere are given in the following pages. Curves for evaluating the standard integrals and the functions appearing in the equations of the flux are reproduced from "The Reactor Shielding Design Manual", edited by Rockwell.(l3) For a more complete discussion Ref. 13 should be consulted. The nomenclature used in these equations is as follows: Scalar flux (cm-2sec-l) So Source strength of point source (sec- ) SL Source strength of line source (cm-lsec-1).SA Source strength of plane source (cm-2sec-l) SV Source strength of volume source (cm-3sec-1) AIs Macroscopic cross section of source material (cm-l) 2, 2,.-.'n Macroscopic cross sections of shields 1, 2,...n(cm-l) ti Thickness of ith shield (cm) n bl Haiti b2 bl + EsZ b3 bl + Ash h Thickness of slab source (cm) Z Effective self-attenuation distance (cm) Ro Radius of disk, cylinder, or sphere (cm) B Symbolic build-up factor n-l e En(b) b e dt for n > O. When n = O, this becomes b -b Eo(b) = 0 b F(Gb) ei _-b sec 9' d.' b G(a,b) Hi F tanl- 1 ab db' where F is the function defined above ab Fn(t,a) e En(b) db 4.32

A number of relationships for sources of various shapes may be derived from Equation 4.11. A. Point Source From Equation 4.11, the flux at P1 in Fig. 4.19 from an istropic point source is 'C. liti 0 = B So e i Ht 4ka2 where Z piti = l1tl + 142t2 + a t3 + s.., is the sum of the individual products i 11 of macroscopic cross section and thickness for the various shields. In the notation given on page 32, this equation may be written S -bl = B O e (4.19) 4 -a2 /. P2 Fig. 4.19. Geometry of point source.(l3) At point P2 in Fig. 4.19, trigonometrical factor must be included in the calculation, and S= B SO -bl sec @ (4.20) 4(a sec )2 4. 33

B. Line Source See Fig. 4.20a SL = Activity/Unit Length For element dU At point P2 -bl sec ( SL(di)e d =L (4.21) 4t(a sec 0)2 P. t P 2For small angles: dx = (a sec @) dO L and dB = dx sec @ di = a(sec @)2 d9 Substituting do SL(a sec @ dg)e-bl sec a 4it(a sec &)2 adG -50. P~~L e-bl sec 0 d( ~- 4ja (4.22) I. d Integrating between limits 91 and 92 LSL ( 1 -bl sec eg + 2 -bl sec d -: 4a e dO + f e d@) 4xa0 0 (4.23) Fig. 4.20a Or in terms of the function F(3,b)(Figs. 4.35-4.41) aometry for element S of line source = B [F(92,bl) + F(@bl)] (4.24) At P1 (see Fig. 4.20b) this equation used with the limits of integration -O1, @p, gives the flux = B -L-[F(92,bl) - F(1,lbl)] (4.25) If P2 is such that @1, = @2, then the integral becomes @1 2 I -bl sec @ d 0 and = B -L F(@l,bl) (4.26) 2ta For no absorber, bl = O and Equation (4.23) becomes 4.34

4xa O SL (9 - 92) (4.27,) / 91 \, P4 Fig. 4.20b Geometry for line source.(.3) For an infinite line source,., = o, and hence @1 2= 82 sO the flux is then given by = B e dO (4.28) B = ~o -Bi seFe( ~SL F(Aj bl) 2ra At P3, for a non-infinite line source, @2 = 0 and 4,55

= B L e-bl sec e -B T. e dO 4aa ' _ = B SL_ _ F(@1,bl) (4.29) 4na For a semi infinite line source, the flux at P3 is given by putting 02 = 0 and. 1 = t/2, and = B F(-7 b1) (4.3o) 4ica At the point P4, for a non-infinite line source, the flux is sLr rg @ -bl sec g = B - [F(l1,bl) - F(@2,bl)] (4.31) 4ta C. Disk Source (K curves) At P1 on the center line of the disk, the contribution to the flux from an elemental of radius r and width dr is SA (2trdr) -bl sec ' (4 d = B. e (4.32) 4Ap2 But p2 = a + r So pde = rdr and sec g' = p/a, so SA dp. a d = B. e (433) 2 p 4.36

drd aP Fig. 4.21. Geometry for disk source.(13) Integrating, making the substitution a t S bl sec G _t f - B. A r e dt 2 t 2bbl t BSA BA [El(bl) - El(bl sec (4.54) At an off-axis point, a similar though considerably more complex derivation gives b, sec - 0 B. - — A Ke te -t d a - 2 b1 t ' Ro 'o BS2 Ro, 45) z iiti where t is a mean absorption coefficient defined by i = K is to be evaluated from the curves given in Figs. 4.23 - 4.26. i t 4.37

4.13 EXAMPLE PROBLEM 2, POINT SOURCE USING METHOD FROM ROCKWELL(13) Example 2: Revaluate Example 1, Part B-4 for a point source using the method from Rockwell, Equation 4.16. Solution: Build-up factor for point source from Fig. 4.16a. A1 = 8.1, A2 = 1 - A1 -7-1 -osl = o.o83, U2 = 00 34 by Equation 4.16 -(ax)2 -(a2X) B = A, e 4 A2 e +(. 083) (6*3) -(oo34)(6.-3) - 8.1 e - 7.1 e +0.522 -0.214 - 8.1 e - 7.1 e = 8.1 (1.686) - 71 (1239) = 13.65 - 5.73 = 7.92 1.239 SO -bl = B( ) e So = 1000(2)(1.25)(3.7 x 101~)(1.6 x 10-6)(3600) = 5.33 x 1011 ergs/hr 5.33(1011)(0.027) 1 72 x 108 /hr 83.8 0=.92 1.72 x 10D8 -6.3 7.92 ( -- r e 't(i) (105)</ 7.92(1.24 x 10)(.00183) 8. O r/hrl or: - iAe + e 4.38

1.72 x 108 e3) - 71 e 4r(lo5)2 = 1.24 x 103 (8.1/340 - 7.1/680) 16.7 r/hr.1 4.14 EXAMPLE PROBLEM 3, LINE SOURCE USING METHOD FROM ROCKWELL(3) A. Using the Rockwell method for a line source calculate the flux at point P in Fig. 4.22 for: 1. A 1000 curie Cobalt-60 line source without 15 in. steel shield. 2. The same line source with the 15 in. steel shield. B. Add 10,000 curies of Gold-198 (which emits a gamma photon of.411 Mev) to the 1000 curies of Cobalt-60 and calculate the flux at "P" for a line source by making an individual calculation for each energy and build-up factor. STEEL E!L 90Cm 15CFig. 4.22. Configuration for example 4.3. Solution: A.1. Line source without shield S =1000(3.7 x 1010)2 = 7.4 x 10ll photons/sec 100 cm 4.39

tan 01 = 50/105 = 0.476 01 = 25.50 = 0.444 radians S F(01,bj) 2rta From Fig. 4.35, for b = 0, 0 = 25.50 F = 0.40 = 7.4(10")(0.4) = 0.448 (109) photons/sec 6.28() (105) cm o.448(1o09) (1.25) (1.6 x 106)(3600)( 27) 7 a 035(103)r/hrl 83.8 A.2. Line source with shield - r1 -(l+acl)[t sec o 92 -(l+O2)kt sec 0 L = A1 e dO + A21 e d| 2xa ' o O - = 5(103)~/,1 -(1.0o83)(6. 3)sec 8d[ - 7.1 e (lc 1.035(10s):8.1 e d8 - 7.1 e dQ 1 035.(103) 5.77 sec 6 sec 1.039(105 ) 8.1 e de - 7.1 e dO o o From Fig. 4.35, for b = 5.77, o = 255.P~, F = 1.02(10-2) b = 6.51, 0 = 25.50" F = 5.0 (10-3) = 1.035(lo)10 8.1(1.02)(10-2) - 7.1(5.0)(10-3) - 8.17 r/hr 4.40

B. Multiple group calculation. J cobalt-60 = 8.17 r/hr, gold-193 Al = 10.2, A2 = - 9.2 al = -0.096, )2 = - 0.011 SL 10,o000(3.7) (o101)(.41)(1.6)(10-6)(36oo00)(0.027) = 2.81(106 o)/ 100 x 83.8 cm From Fig. 4.6 = 0.09, Atx =.09(7.78)(15) = 10.48 2.81(106) 10.2 G-(1-el096+)(1048)sec Gd 2f2 (l+.011)(10.98)sec 9d 2'( ) 10.2 d~G - 9 d. 27(10s) 0 - 25e 5 -9.48 sece 255 -Lo.60 sec 9 4.27(103) 1 1 ~0.2 i~e cd - 9.2 e d From Fig. 4.36, for b = 9.48, 9 = 25.5, F = 2.2(10-5) b = 10.60, 9 = 25-5, F = 0.80(10-5) = 4.27(10l) 10.2(2.2)(10-5) - 9.2(0.80)(10-5)i 4.27(10 s)(15.08)(10-) = 0.642 r/hr total = 8a7 + 0.64 =8.81 r/hr 4.41

1.0 0 4 I.8 ~~OZ~~~~~~d/Ro -F' o 0 Fig. 4.23 K curves for disk source1.6 1. 2.0 Fig. 4.3 K curves for disk source off center line with a/Rnd 7. (13)0 for u Ro parameters of 1, 4, and 7. (13)

o/Ro =0.25 K vs. d/Ro forFjRo =,4,ond 7 1.0 rF 4 0 0 6.2 0.2.4. 6.8 1.0 1.2 1.4 1.6 1.8 2.0 d/Ro Fig. 4.24. Disk source off center line.(13)

a / Ro = 3.0 K vs d/Ro for F Ro= 1,4,and 7 1.0 nd.8 " 0 Ir 4 10 0.2.4.6.8 1.0 1.2 1.4 1.6 1.8 2.0 d/Ro Fig. 4.25. Disk source off center line.(13)

a/Ro=5.0 K vs d/Rofor Ro = 1,4, and 7 1.0 OL.8 0 4:-.4.2 0~.2.4.6.8 1.0 1.2 1.4 1.6 i.8 2.0 d/Ro Fig. 4.26. Disk source off center line.(13)

4.15 INFINITE SLAB SOURCE (1) (a) For exterior points. This case may be considered by putting @ = rT/2 in the case of the disk source with Ro = co. This process gives in the notation given on page 32, when SV is considered as a function of x and kix SV(X) = Si e, P LZB S kib3/ks -k. B 2 S. ekib(b3, -_) _ Fl(bl -k)! (4.36) If the slab is infinitely thin, h = o and S n -t = B A. e dt (4.37) 2 tb ol X Fig. 4.27. Geometry of an infinite slab source and exterior point. (b) Interior points. If, as before SV is a function of x and is Sv(x) = ie i the flux is given by i 0 = 2PkB i Fi(ksd, -k) + F1 s(h-d) ki3 (4.38) 2Cls ~~~~i P-I~-S 4. 46

d l. —| dx X (13) Fig. 4.28. Geometry of infinite slab source and interior point.(13) Fl(t,a) which appears in Equations 4.36 and 4.38 may be evaluated from the curves given in Figures 4.29 - 4.33. Ii.. 4

~~2.0 _____a = 403.6 3.2 2.8 2.4 2.0 2.0 '~~~~~~~. 1.6 1.8 1.4 1.6 6 1.2 1.4 / 0 1.0.8.6.6.4 1.2 0 -. 8 -.2 -3.6 ~~~-, -1.0 '.2: - " ~~1.2.60- -2 1.4 -1.6 -1.8,m -2.0 ~~.4 ----— fT-~ ~ --- —2.4 -2.8 -3.2 3-.6 - 4.0.2 F,=O to 2; t =0 to 1.4; a=-4 to4 0, 0.2.4.6.8 1.0 1.2 1.4 t Fig. 4.29. The function Fl(t,a).(13) 4. 48

2.2 o= 4/.02.4 2.01.8 1.6 1.4 1.3 1.2 4. 0O, 1.15 3.6 1.10 3.2// 4.05 2.8, 2.4 2.0.8 1.6 1.2.8 2 --1.2 -1.4 F =0 to4; t=Oto7; a=-4 to4 1 0 I,, t -3.2

3.8 3.4 3.0 104 a= 4.03.6/3.2/ 2.8 2.6 2.4 2.2 2.0 - | 1 il It I I ~.8 103 I11A t/ I _ _ III/ // 1.8 I 10m1;~~~~~~~~~~~~~~~~~~~~~~~~. 1~~~~~~~0~~7 103 " " ' 11 I I 1.0 10ode I 4. lF1=0 to I0;t-0to14;a=-4 to4 ~Fig. 4.5.1. The function F1(6ta).(13) 0450 F I to 10; t=Oto 14; a -4 to 4I1 T ill. l.1 Ih I!cto / / I / ~.(U i IIIIIII I il! / ~4,i //

09 a = 40 3.8 3.6 3.4 3.2 3.0 2.8 2.6 Im ~~ ~~ ~~ ~~~~~~~~~ I I I I1 / 1 I1 F1 =14to 109 t= 0to 14;a=2to4 108 2.4 8 - ~~~~~~I iI I I I I / ~~~~~~I, I! 111' /1!1 ~~~~I i I 111 1!/ I /d. 11 I I i I 107! O E E ZIIIiI/ 2.0 1041 60 2 4 6 8Il 1 / /2 14 I! I I I /........1,,11/1'I // X III I j / / /' /2 Fig. 4.32. The function Fl(t,a).(13) 4,51

1014 a=4.0 3.8 3.6 10 - I F- 109 to i014; t0=to 14; a=2.8 to4 3.4 1012 3.2 10 I 12 109 - i... 52

4.16 CYLINDRICAL SOURCE(13) Exterior on Side (EsZ Curves) / / Ro~*-\C'~ I \ // I / hlux sii I y.I P BSvR2 = V~[F(llb2) + F(Q2,b2)] (4.39) 4(a+z) where 91 is different from 92. At P2, if A1 = 2 = 0t BSVRS F(G,b2) (4.40) 2 (a+z) If h = oo, i.e., it is an infinite cylinder, then 9 = r/2. At P3 SvR [F(@2,b2) - F(@l,b2)] (4.41) 4(a+z) 4.53

Function F(G,b), plotted for d ifferent arguments of G and b, is given in Figs. 4.35 - 4.40o. Figure 4.41 gives the curve for self-absorption distance Z for a cylinder with a/Ro > 10. Curves for the case of a/Ro 4 10 are given in Figs. 4.42 and 4.43. 4.54

9O0~~~ Fvs b 90 ~ 10 60~ For F=10-4to 10;b-0 to7;8=1~ to 900 \ 340 %\- _300 o 20 _ 0~ 11 00 100 10;-,o- \\ \\:> 8~ % __ I I I I% X_ ",_, 0 2 3 4 5 6 7 Fig. v.55. Txe function ' '. 4.55

-2 900 400 i0. -5 105'63 0 N5 ~ 100 x........ x NX.... 56 i X ']\, %. '\.%. \\%.%. '...... % \~ 5 6 7 8 9 10 11 12 b 4.56

90~ 200 150 100 Fvs b 66| For F =lO1IOtolO- 5 b 11 to 18;= 1~ to 90~ i0 IIl 12 13 14 15 16 17 18 Fig. 4.37. The function F(G,b).(13) 4.57

10 900:.~._9Q0 I 150 10 __K___s '20Fv I6 ~ 15 100 For fl1612to I0 b16 to 2;8=I to90 0 80 1I0-80 10 - 9 I l I \" I0 10 16 17 18 19 20 21 22 23 b Fig. 4.58. The function F(3,b).(3) 4,.S$

-9 10 90 -10 2 A 5~ For F=1l to 10;b=20 to27;8=1~to90~ 1511 -D 11 -, - 20 21 22 23 24 25 26to 10 Fig. 4.59. The function F(\! ).(13) 4.59 iC14 20 21 2? 23 24 25 26 27 4-.59

09 *t' (c)'* ( q'e), uoi.ounj aG ~l o'0t ' *T.rI q ZE I~ 0~ 6Z 83 LZ 9Z 9z 91 -01.o 9191-01......... I(,, x x x x, 0 ~~~~X ~~~~~~~~~~~~~.~~..... ~ 0......... -; I -: 01

5.0 4.0 2.0 1.0 O, 0 2 4 6 8 10 12 14 16 18 /J sRo Fig. 4.41. Self-absorption distance, Z, for cylinder with a/Ro L 10.(13) 4.,61

<V(a+ Rs O 1 24 a 20 2.2 f ora/ Ro< 10-I _ 14El 8,2""~ 1. 10 0.6 ___ 0.4 _ _ ___-_ - 1 8.5 1.2 _ -_" 1.0 7! 0.6 4 0.4 0.2 - ~ 4.6 Note: Knowing Ro,a,and,find m. Use m in next graph to obtain usZ 0 I 2 3 4 5 6 7 8 9 a/Ro 4.62

3.6 a/Ro= 3.4 3.2 ~~~2.86~ 2.41 /1 / "/,. 2.4 3(.~~ ~~0 2.28_~ - 10 I4 2.o Y l/ ///// 241 eO00 00 2.0 1.6~ i.2 -1.0, 0.8 0.6 9~I I 1 1 0.4 Note' Knowing a/R and b, find (I/m)L, Z, 0.2 multiplying by m (from previous graph) to obtain usZ. 0~~~ i 1 1 i- I I I I I I 0 2 4 6 8 I0 12 14 16 18 20 22 24 bl Fig 4~~~~~~~.63 efasrto isacZ faclndra ucin

Exterior on End (h < 3/h,): DTe problem can be dealt with by considering two truncated cones with half angle 01 and 02o This gives the upper and lower limits for the flux. 1. Upper limit At P1 in Fig. 4.44 BS= B E2(bl) - E2(b3) + E2(b3 sec -1) E2(bl sec 01) (4.42) 2~ts!sec 01 sec 01 - - / Ok I \\ / ~~At PI / I / I / I, / I h \ /I \ I,\ Fig. 4.44. Cylinder. 2. Lower limit At P1 - Et(b1) - E2(b3) + E2(b3 sec 92) _ E2(bl sec 02) (4.4) sec G2 sec G2 These are rather wide limits which can be reduced if h ~ 3/C.s. In this case the flux may be given the following limits: 4, 64

1. Upper limit At P1 BSVV()b E2 (b, sec_ (1) (4.44) = VBSv E2(bl) _ E2(bl sec O!) 2is' sec - 2. Lower limit At P1 _= BSV E2(bl) E2(bl sec 03) (445) 2Ls, Lr sec 03 where 93 is given by 93 = tan-l Ro (a+h') and (4.46) A _s Interior (G Curves) For this case, the results will be given briefly to avoid lengthy mathematics. If P1 and P2 are on axial center line, the flux at P1 is given as 0= Sv [G(%shj,b) + G(tsh2,b)] (4.47) 2is where b =- AsRo. And at P2 - G (%sh,b) (4.48) 2ps where b - sRo. 4 65

h Fig. 4.4. Cylinder.( Fig. 4.45. Cylinder.(13 ) At P3, only the limits can again be stated, and the upper limit of the flux is: Sv [G(pshl,b5) + G([sh2,bs) + G(Lshl,b6) + G(tSh2,b6) ] (4.49) 4ts At P3, the lower limit is BS= V [G(Cshl,b6) + G(llsh2,b6) + G(itShl,b4) + G(ItSh2,b4)] (4.50) where b4 - ks(Ro-d), b5 - is(Ro+d), and b6 - rs R=2o-d2 Function G(p.sh,b) for different values of the argument Atsh may be obtained by referring to Figs. 4.46 and 4.47. 4.17 SPHERICAL SOURCE(13) (a) Interior points. The mathematics is again complicated for this cas and only the results are quoted. The flux at P1 is given by,= BL( e-sR o) (4.51) as

b 1.0 5.0 -I oo 3.0 _ _ ~2.0.9 1.5.8 1.0.7 o8.6 G I, l0.6.5 0.4 467...... 0.3.2 0.1 Gvs p.sh for G=0 toI-,s sh=0to3.5; b=0.I to co 0 1 2 3 4.67

1.0 1.5 0.9 1.0 0.8 0.7 / 0.2 0.6 0.5 0.2 0.4 s3 h =0.1 0. i 0.2 Gvs. b l. For G =0 to I;/ish =0.1 to Io; b =0 to 7 0. 0 1 2 3 4 5 6 7 Fig. 4.47. The function G(rshb).(l3) 4 68

Fig. 4.48. Sphere. (13) And at P2, BS -VR e +..... (1 1 + e-2sRo! (4.52) 2ts V 2tsRo 2,sRo, At P3 BSV eb4 -b 1 b-b b [e 1b4) e.(1+b5)] + [Ee-(b4) -E(b5)]+ [is 2 2 4Cisd 4Ctsd (4.53) where b4 - s(Ro-d) and b5s - s(Ro+d). (b) Exterior points (Z curves). --- I —1 a<49 Got prasuel a Fig. 4.49. Geometry of spherical source.(15)

At P1 = 2/3 BSvRo[El(b2) - El(b2 sec ) ] (4.54) At P2, if bl = 0, -2 tsR0~., = BSv' 1 +BSV '" (4.-55) 2~ts 24sRo 21sRo / Figures 4o50 and 4.51 show curves for the self-absor ti9n distance Z for sphere with a/R0 >1 and for a/Ro < 1, respectively.(l3) 4E 70

.70.65 5.0.60 45.55 4.0.50 3.5.45 3.0 40 Q ~~~~~~~~~~~~~~~~~~~~~0 r~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~C H N N 2.5.35 2.0 30 1.5) /1 I I I I I I I 1.25 1.0.20 0.5.15 0 4 8 12 16 20 24 28 32 36.10 2 4 6 8 10 12 14 16 18 1u8 R0,us(a + Ro) Fig. 4.50. Self-absorption distance Z for Fig. 4.51. Ratio of self-absorption distance sphere with a/R0 =1. (13) to radius of sphere, Z/RO, for a/R- 1(

4.18 TYPES OF GAMMA RADIATION SOURCES Large-scale operation of research and power reactors will make available gamma-radiation sources of strengths varying from millicuries to megacuries. The majority of these isotopes emit one, two, or three gamma rays with welldefined energies and the calculation of shielding for such sources is consequently easier than for complex-spectrum sources. However, the gross fission-product waste produced by an operating reactor has an extremely complex spectrum and yield. In the class of well defined gamma energies,cobalt-60 and caesium-357 sources merit consideration while the fission gases and spent fuel elements are examples of complex-spectrum sources. The whole spectrum may be divided into a set of discrete energy groups whose source strengths have been estimated. The problem is, therefore, reduced to the determination of the total intensity due to the discrete set. Typical of these complex spectrum sources is a spent MTR fuel element. 4.19 MTR FUEL ELEMENTS Spent MTR-type fuel elements may be made available in large quantity as sources of gamma radiation for industry. Transportation and installation of these sources require detail shielding considerations. The MTR fuel element may be considered to emit seven groups of gamma rays of energies 0.35, 0.5, 0.7, 0.9, 1.25, 1.75, and 2.5 Mev. Extensive theoretical and experimental studies of the gamma-decay dose rate and heating from the spent MTR fuel elements of different histories had been reported by Francis and Marsden.(l4) Figure 4.52 shows a sketch of an MTR fuel element as described in the above report. Fig. 4.52. Sketch of MTR type fuel element. (15) 4.72

4.20 EXAMPLE OF GAMMA SHIELDING CALCULATION FOR MTR FUEL ELEMENT[ The calculation of the radiation field and the minimum shield thickness required to reduce the dose rate to the desired tolerance level can be carried out in the following steps. 1. Type of Radiation Source. —An MTR fuel element is used in a 30 -megawatt reactor for 17 days and the spent element is then cooled for 30 days. The source is surrounded by 2 inches of water. 2. Nature of Nuclear Radiations.-Gamma radiation. 3. Sources and Energy Spectrum.-QV is the source strength in photons/sec at equilibrium and E is the energy of the different energy groups of gamma radiation, tabulated for one fission/sec (see Table 4.7). To obtain the total energy per sec (Mev/sec), S should be multiplied by the total number of fissions/sec, obtained from the irradiation history. 4. Type of Shield.-For transporting a fuel element, the shield used should be as light and compact and preferably as economical as possible. From TABLE 4.7 ENERGY SPECTRUM FOR MTR FUEL ELEMENT(14) Energy-E Source Strength-QV QE = S No. (Mev) (photons/sec) (Mev/sec) 1 2.5 2.6 x 10-4 6.5 x 104 2 1.75 5.99 x 10-3 1.05 x 10-2 3 1.25 2.6 x 10-5 3.25 x 10-5 4 0.9 2.72 x 10'3 -2.45 x 10-3 5 0.7 7.9 x 10-3 5.53 x 10-3 6 0.5 5.98 x 10-3 2.99 x 10 -7 0.35 8.5 x 10-3 2.98 x 10-3 the point of view of compactness, lead is to be preferred. In addition to the lead, the two inches of cooling water and two inches of structural steel may also be considered to be attenuating the gamma radiation. In addition to these external attenuation media, self-absorption due to the source itself may be considered. 4. 73

5. Attenuation Processes. —Gamma-radiation intensity is affected by spatial attenuation, exponential attenuation, and the build-up factor due to scattering. Since the heavier material (lead) follows the lighter materials (water and steel), the build-up factor for lead is used. For steel and water then, only exponential attenuation is considered. 6. Nuclear Constants. —The linear absorption coefficients (. cm-l) for energy groups for water, source material, steel, and lead and the energyabsorption coefficient (-oas/p)(cm2/gm) for tissue are given in Table 4.8) TABLE 4.8 ABSORPTION COEFFICIENTS FOR VARIOUS ENERGY GROUPS OF MTR ELEMI T(14) No. E= (~.-5 Nev) (1.75) (1.25) (0.9) (0.7) (0o.5) (0.35) 1 kH20 0.044 0.054 o0.0o64 0.075 0.085 0.098 0.113 2 clsource 0.067 0.083 0.097 0.112 0.125 0.146 O.169 3 ['steel 0.290 0.328 0.399 0.485 0.54Q o.635 0.767 4 ~Ilead 0.485 0.54 0.66 0.86 1.1 1.65 3.0o8 5 (1-as/P)tissue 0.0225 0.0255 0.028 0.0305 0.0315 0.032 0.032 6 Z, 2.69 2.69 2.68 2.59 2.56 2.47 2.43 self-absorption distance, cm Besides these nuclear constants, the dose build-up factor coefficients for lead are given in Table 4.9 for these energies for a point source in an infinite homogeneous medium. TABLE 4.9 DOSE BUILD-UP FACTOR COEFFICIENTS( 12,13) No. E = (2.5 M&v) (1.75) (1.25) (0.9) (0.7) (0.5) (0.35) 1 Al 2.3 2.75 2.6 2.5 2.3 2.2 2.1 2 A2 -1.3 -1.75 -1.6 -1.5 -1.3 -1.2 -1.2 3 al -o.65 -0.05 -0.04 -0.03 -0.02 -0.015 -0.01 4 ~2 0.125 0.135 0.14 0.14 o.14 0.14 0.14 7. Source Geometry.-For the purpose of calculation, the MTR fuel element may be considered as a cylinder of 2 ft effective length and 3 inch diameter. Intensity at any point can be determined by referring to Equation 4.39. However, for the purpose of illustration, the intensity at an exterior point P on the midplane perpendicular to the source as indicated in Fig. 4.55, is discussed, where 4.74

h = height of cylinder = 61 cm, Ro = radius of cylinder = 3.81 cm, r - distance between the surface of source and the field point P, = (5.08 cm of water + 5.08 cm of steel + t cm of lead + t' cm of air), Z = effective self-absorption distance (it depends on geometry of the source and the energy of the radiation), and 0 = tan-l h 2(r+Z) 8. Total Intensity, (I). —The total intensity is the sum of the intensities contributed by the seven energy groups at the point P. 7 I = Z Ik (r/hr (4.56) k=l where Ik = intensity for k-th group. 9. Intensity for k-th Group, Ik. —Intensity Ik for k-th gamma-ray group is the product of Iok intensity at P due to spatial attenuation, Bk = buildup factor, and Fk(G,b) = exponential attenuation. Ik = Iok - Bk ~ Fk(G,b) r/hr (4.57) As stated earlier, the build-up factor Bk, if expressed as the sum of two exponentials, as shown in Equation 4.58 can be incorporated in the exponential attenuation factor Rn ~ STEEL STEEL LEAD Fig. 4.53- Geometry for example calculation. 4, 75

Fk(G,b), i.e., Bk = A1 e 1t + A2 e-C2 t (4.58) The product of Bk and Fk(G,b) can then be replaced by Gk(G,b) i.e., Gk(G,b) = Bk, Fk(G,b) (4-59a).. Ik Iok ~ Gk(G,b) r/hr (4.59b) a. Determination of Gk. n -bn sec ' Gk = Z An e d.' n=l o As pointed out by Taylor,(l2) the use of only two exponential terms in the build-up factor is sufficient for most calculations. This simplifies the above expression(if build-up is considered significant only in the lead)to: G /A r [[LsZ + 5 08 ktH20 + 5 08 1lFe + (1 + cz1) ~Pb t]sec G' o j (4.60) All the quantities in Equation 4.60 for Gk are known. Therefore, Gk can be calculated by referring to Figs. 4.54, 4.55, and 4.56 for the evaluation of the above integral F(9,b2). b. Determination of Iok. Gok = (0ok)(Ak) P r/hr. (4.61) 4. 76

To determine Iok' the factors ooky hkk and P must be evaluated individually. (1) Ook (Energy Flux) 0ok = energy flux (Mev/cm2 ~sec) According to Equation 4.40 aok }Sv Ro. Mev 2 (r+z) Lcm2a secj where Sv = source strength (Mev/cm3.sec) or Sv = S/volume = QVE/TRo2 h Converting the units of time from seconds to hours, S Mev (3.6 x o03) Mev. (4.62) 2ch(r+Z) cm2.hr/ k jok can be evaluated since for the different energy groups S and Z are known. (2) kk (Conversion Factor) kk = conversion factor to obtain dose rate in r/hr. gm of air! c1.6 x 10-6 ergs () L cm 83 ergs Mev P(tssue) g iI i i ik(tissue) L r 2 k = 7 --- (1.93 x 10-4) (r) (cm2/Mev). (4.63) P k k(tissue) This is based on the fact that one roentgen is equivalent to the absorption of 83 ergs of gamma radiation by one gram of standard air. 4.77

~(91) 0(P0/ _I= (qj o)j jo UOiTenTeAe *i5Q-.BTi b~ ~~ Z ~ 1~ 0~ 6Z 8Z ILI - 01 91 -01 o o, % 01 ~1 -l ll ll l l I. I I I I~~~~~~~~~~~~~~~~~~,_

6L 1t 0. ( ~9T) astq J = (q' ),i am U To uo0nT'A, '5't 'e. _.q It Oft 62 8~ 1~ 9~ S.... CN~_-1~~~~~~~~~~01 II101....00 01 91 -\WKI \ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~-llllI I XX II o O -~ I I\ I \ I I I > \ ' \.~~~~~~~~~g ~~~\'\ '..t >06 \ \ oC; -~~~~~9,P.....\LI... L Xo\

I01 10 l 0 go -21 I0 -22 I0 -23 I0 42 43 44 45 46 47 48 49 Fig. 4.56. Evaluation of F(~,b) = f ebsec~d~.(l6) 4,. 80

([Pdt J8 X= (i'o)j Jo uoTnTenrT L-tA * 0 tI 9c C S bc ic 2c I 05 6t0 01 01 01 01

(3) P (Total Fissions/Sec) P is the total number of fissions in 1 sec and is given by p AM x 6.02 x 103 (fissions/sec) (4.64) 235 T where AM = burnup of the fuel element (gmn/sec) and T = irradiation time in sec (MWD/MW x 86400) where 1 day = 86400 sec MWD = megawatt days MW = power rating of the nuclear reactor in megawatts. S Mev Iok = x2l+) (3.6 (1.9 x 10-4) 2Jh(r+Z)>jk hr k(tissue) I J ~ 1i6 AM(gin) *i(r) cm2, [2.97 x 118] TM(gm) Mev ' T(days) or Iok - r+ k L AM (gm) e (2.06 x 1016) r/hr 2jth~ P 1T (d~ays)(4.6)) ok!2h(r+Z)'k p lk(tissue) (ay (465) When Iok and Gk for the k-th energy group have been determined the total intensity due to the whole spectrum can be obtained by summing up the contributions due to the individual groups. For shielding purposes, only two groups (2.5 Mev and 1.75 Mev) need be considered. Thus summation of intensity for k = 1 and k = 2 is sufficient. For additional information on gamma shielding calculations the reader is referred to References 17-72. 4,82

REFERENCES - CHAPTER 4 1. Brownell, L. E., "Radiation Uses in Industry and Science," USAEC, OTI, June, 1961. 2. Goodman, C., "The Science and Engineering of Nuclear Power," Addison-Wesley, 1952. 3. White, G. R., "Gamma Ray Attenuation Coefficients from 10 Kev to 100 Mev," NBS-1003, May 13, 1952, Revised version published as NBS Circular 583, April 30, 1957. 4. Kinsman, S., "Radiological Health Handbook," Sanitary Engineering Center, Cincinnati, Ohio, 1954. 5. Lovewell, P., "Industrial Uses of Radioactive Fission Products," Stanford Research Institute Report, 1951. 6. Siri, W., "Isotopic Tracers and Nuclear Radiations," McGraw Hill, 1949. 7. Goldstein, H., and Wilkins, J. E., Jr., "Interim Report on the NDA-NBS Calculations of Gamma Ray Penetration," Memo 15C-20, Nuclear Development Associates, Inc., White Plains, N. Y., Sept., 1953. 8. Braestrup, C., et al., "Protection Against Radiations from Radium, Cobalt-60 and Cesium-137," Nat. Bur. of Standards Handbook 54, 1954. 9. Marinelli, L. D., Quimby, E. H., and Hines, G. J., "Dosage Determination with Radioactive Isotopes, II —Practical considerations in therapy and protection." Amer. J. Roent., 59, 260, 1948. 10. Goldstein, H., "The Attenuation of Gamma Rays and Neutrons in Reactor Shields,tt USAEC Contract AT(30-1)-862, U. S. Govmt. Print. Office, Wash. 25, D. C., May 1, 1957. 11. Spencer, L. V., and Fano, U., "Penetration and Diffusion of X-Rays; Calculation of Spatial Distributions by Polynomial Expansion," J. Research Natl. Bur. Stds., 46, 446, 1951; Phys. Rev., 81, 464L, 1951. 12. Taylor, J. J., "Application of Gamma Ray Build-Up Data to Shield Design," AEC publication, WAPD-RM-217, Jan., 1954. 13. Rockwell, Theodore, III, t"Reactor Shielding Design Manual," Rept. TID-7004, Office of Technical Services, Washington 25, D. C., March, 1956. 14. Francis, W. C., and Marsden, L. L., "Experimental and Theoretical Values of the Gamma Decay Dose Rate and Heating from Spent MI'R Fuel Elements,t" AEC Research and Development, Rept. IDO-16247, Jan., 1956. 4.83

REFERENCES - CHAPTER 4 (Continued) 15. Fahnoe, F., et al., "MTR Rods as Fission Product Sources for Industrial Sterilization," KLX-1395, Vitro Corp., Jan. 2, 1954. 16. Dennis, R., Purohit, S. N., and. Brownell, L. E., "Procedures for Shielding Calculations," Rept. 1943:8-86T, Eng. Res. Inst., The University of Michigan, Ann Arbor, Mich., Jan., 1957. 17. Gamble, R. L., "Fission Gamma-Ray Spectrum," Rept. 1620, Oak Ridge National Laboratory, 1953, p. 15. 18. Francis, J. E., and Gamble, R. L., "Prompt Fission Gamma Rays," Rept. 1879, Oak Ridge National Laboratory, 1955. 19. Moteff, J., "Fission Product Decay Gamma Energy Spectrum," APEX 134, General Electric Company, 1953. 20. Clark, F. H., "Decay of Fission Product Gammas," Rept. 27-39, Nuclear Development Associates, Inc., 1954. 21. Mittleman, P. S., and Liedtke, R. A., "Gamma Rays from Thermal-Neutron Capture, Nucleonics, 13(5), 50, 1955. 22. Stephenson, R., "Introduction to Nuclear Engineering," McGraw-Hill Book Co., N. Y., 1954. 23. Preiser, S., Mittelman, P. S. and Berndtson, C. R., Plane Isotropic Gamma Ray Buildup Factors in Lead and Water with Applications to Shielding Calculations, NDA 10-144 (Dec.21, 1954). 24. Blizard, E. P., Paper presented at the Nuclear Engineering Congress at Cleveland, Ohio, Dec., 1955. 25. Moteff, J., "Miscellaneous Data for Shielding Calculations," APEX 176, General Electric Company, Dec. 1, 1954. 26. Blizard, E. P., "Nuclear Radiation Shielding," Annual Review of Nuclear Science, 1955. 27. Way, K., and Wigner, E. P., "The Rate of Decay of Fission Products," Phys. Rev., 73, 1318, 1948. 28. Ajzenberg, F., and Lauritsen, T., "Energy Levels of Light Nuclei," Revs. Mod. Phys., 24, 321, 1952. 29. Fano, U., "Gamma-Ray Attenuation," Reactor Handbook, Technical Information Service; U. S. Atomic Energy Commission Document, AECD-3645, Aug., 1955. 4. 84

REFERENCES - CHAPTER 4 (Continued) 30. Fano, U., "Gamma Ray Attenuation I," Nucleonics, 11(8),8, 1953. 31. Fano, U., "Gamma Ray Attenuation II," Nucleonics, 11(9), 55, 1953. 32. Heitler, W., "Quantum Theory of Radiation," 3d ed., Oxford University Press, New York, 1954. 33. Bethe, H. A. and Ashkin, J., in E. Segre (Ed.), "Experimental Nuclear Physics," Vol. 1, pp. 305, 349, John Wiley and Sons, Inc., New York 1953. 34. Spring, K. H., "Photons and Electrons," John Wiley and Sons, Inc., New York, 1950. 35. Hughes, A. L. and Dubridge,, L. A., "Photoelectric Phenomena," p. 193, McGraw-Hill Book Company, Inc., New York, 1932. 36. Nelms, Anna T., Graphs of the Compton Energy Angle Relationship and the Klein Nishina Formula from 10 kev to 500 Mev, NBS-542 (Aug. 1953). 37. Latter R. and Kahn, H., Gamma Ray Absorption Coefficients, R-170 (Sept. 19, 1949). 38. Goldstein, H., Estimates of the Effect of Fluorescence and Annibhilation Radiation on Gamma Ray Penetration, NDA 15C-31 (Feb. 26, 1954). 39. White, W. E., High Energy Gamma Ray Penetration in Lead, NEPA-1324 (Mar. 6, 1950). 40. Snyder, W. S. and Powell, J. L., Absorption of Gamma Rays, ORNL-421 (Mar. 14, 1950). 41. Goldstein, H. and Wilkins, Jr., J. E., Calculations of the Penetration of Gamma Rays, NYO-3075 (June 30, 1954). 42. Placzek, G., The Functions of En(x) = e-xu u-n du, Chalk River Report MT1 (NRC 1547), Nat. Research Council Can. (Nov. 1946); reprinted in Tables of Functions and of Zeros of Functions, Applied Mathematics Series, No. 37, p. 57 f., Nat. Bur. of Standards (1954). 43. Goldstein, H. and Aronson, R., Status Report on Calculations of Gamma Ray Penetration, NYO-3079, NDA 15C-1 (Aug. 20, 1953). 44. Rossi, B., "High Energy Particles," p. 228 f., Prentice-Hall, New York 1952. 4.85

REFERENCES - CHAPTER 4 (Continued) 45. Welton, T. A., A Review of Analytical Methods for the Calculation of Neutron and Gamma Ray Attenuations, TID-256 (Nov. 15, 1955). 46. Young, G., Piece-wise Grueling Solutions for Hydrogen, ORNL-415 (Sept. 28, 1949); also, On Straight Ahead Gamma Transmission with A Minimum in the Cross Section, ORNL-416 (Sept. 26, 1949). 47. Bethe, H. A., Fano, U. and Karr, P. R., Penetration and Diffusion of Hard X-rays through Thick Barriers. I. The Approach to Spectral Equilibrium, Phys. Rev., 76:538 (Aug. 15, 1949). 48. Wilkins, Jr., J. E., Openheim, A. and Solon, L., The Transport Equation in the Straight Ahead Case, NYO-633 (Sept. 1, 1950). 49. Solon, L. R. and Wilkins, Jr., J E., Straight Ahead and Root Mean Square Angle Calculations for 20 mc2 Gamma Rays in Lead, NYO-635 (Dec. 15, 1950). 50. Solon, L. R., Wilkins, Jr., J. E., Oppenheim, A. and Goldstein, H., Gamma Transmission in Iron, Tungsten, Lead, Uranium and a Pure Compton Scatterer by Root Mean Square Angle Calculation, NYO-637 (Apr. 5, 1951). 51. Cave, L., Corner, J. and Liston, R.H.A., The Scattering of Gamma Rays in Extended Media, I: Perpendicular Incidence on a Plane Slab, Proc. Roy. Soc. A, 204: 223 (Dec. 7, 1950). 52. Corner, J., and Liston, R.H.A., The Scattering of Gamma Rays in Extended Media,II: Back-Scattering of Gamma Rays from a Thick Slab, Proc. Roy. Soc. A, 204:323 (Dec. 22, 1950). 53. Peebles, G. H. and Plesset, M. S., Transmission of Gamma Rays Through Large Thicknesses of Heavy Materials, P-155 (June 9, 1950). 54. Peebles, G H., Gamma Ray Transmission Through Finite Slabs, Part I, AECD3239, RM-653 Pt. I (July 23, 1951); and Part II, RM-653 Pt. II (May 2, 1952). 55. Peebles, G. H., Gamma Ray Transmission Through Finite Slabs, R-240 (Dec. 1, 1952). 56. Whittaker, E. T. and Watson, G. N., "A Course of Modern Analysis," Chap. 15, Cambridge University Press, New York, 1946. 57. Erdelyi, A., et al., "Higher Transcendental Functions," Vol. 2, Chap. 10, McGraw-Hill Book Company, Inc., New York, 1953. 58. Berger, M. J. and Doggett, J. A., Gamma Radiation in Air Due to Cloud or Ground Contamination, NBS-2224 (June 1, 19553). 4.86

REFERENCES - CHAPTER 4 (Continued) 59. Berger, M. J., Penetration of Obliquely Incident Gamma Rays, unpublished NBS Report (1955). See also J. Research Nat. Bur. Standards, 56:111 (1956). 60. Wilkins, Jr., J. E., Singly Scattered Angular Flux of Gam a Rays at the Source Energy and at the Single Scattering Cutoff, NYO-6273, NDA-15C-46 (Feb. 10, 1955). 61. Certaine, J., Angular Distribution of Photons from Plane Monoenergetic Sources, NYO, 3074, NDA 15C-10 (June 1, 1953) 62. Householder, A S., "Principles of Numerical Analysis," pp. 242-246, McGrawHill Book Company, Inc., New York, 1953. 63. Meyer, H. A. (ed.), "Symposium on Monte Carlo Methods," John Wiley and Sons, Inc., New York, 1956. In succeeding references to this volume it will be labelled SMCM. 64. Goertzel, G., Quota Sampling and Importance Functions in Stochastic Solution of Particle Problems, ORNL 434 (June 21, 1949). See also H. Kahn, Modification of the Monte Carlo Method, P-123 (Rand) (Nov. 14, 1949), and article by M. Kalos and G. Goertzel to appear in Vol 2, Series I, "Progress in Nuclear Energy." 65. Kahn, H., Stochastic (Monte Carlo) Attenuation Analysis, Rand P-88 (rev.) (July 14, 1949); and quoted in T. A. Welton, A Review of Analytical Methods for the Calculation of Neutron and Gamma Ray Attenuations, TID-256 (Oct. 5, 1949). 66. Shor, S. W., V. Computation of Radiation Shield Thickness by the Monte Carlo Method, MIT Tech. Report No. 32 (Jan. 1950). 67. Hayward, E. and Hubbell, J. H., The Albedo of Various Materials for 1 Mev Photons, NBS-2768 (Sept. 11, 1953); also Phys. Rev., 93;955 (1954). 68. Preliminary Report in M. H. Kalos, A Monte Carlo Calculation of the Transport of Gamma Rays, NDA 56-7 (July 31, 1956). 69. Carlson, B., The Monte Carlo Method Applied to a Problem in Gamma Ray Diffusion, LADC 1633, AECU-2857, 1953. 70. Ogievetskii, I., The Theory of Propagation of Gamma Rays through Matter, Soviet Physics JETP (New York), 2(2):312 (Mar. 1956); and Angular Distribution of Gamma Rays at Great Depths of Penetration in Matter, ibid., 319 (Mar. 1956). 4.87

REFERENCES - CHAPTER 4 (Concluded) 71. Aronson, R., Some Remarks on Source Geometry and Single Scattering of Gamma Rays, NDA Memo 15C-37 (May 1, 1954). 72. Taylor, J. J., Applications of Gamma Ray Buildup Data to Shield Design, WAPD Memo RM-217 (Jan 25, 1954). See also Chap. 9, Part V of Reactor Shielding Design Manual, TID-7004 (Mar. 1956). 4. 88

Chapter 5 Counting Nuclear Radiations A number of problems of radioanalysis depend on the accurate determination of the disintegration rate of a radioactive sample. This is true for "tracer" techniques, determination of half-life for radioisotope identification, and various other assay methods. For a given radioisotope the rate of disintegration is directly proportional to the rate of emission of nuclear radiations. The "counting" of the nuclear radiations over a given period of time permits determination of the rate of emission. Today, electronic instruments known as "counters" are used for this procedure. However, early workers counted alpha particles by the tedious process of observing with a microscope individual scintillations on a plate coated with a phosphor and counting the events by eye. The equipment required for rapid and accurate counting consists of a detector such as a G-M tube, scintillation counter, etc., and auxiliary equipment for shielding the sample and timing and recording the counts. This chapter describes first the auxiliary equipment of shielding and electronic counting devices. The theory of counting randomly occurring events is discussed in detail: consideration is given to standard errors and systematic corrections for dead-time, geometry, tube efficiency, scattering, absorption etc. The chapter ends with paragraphs on the major problems associated with counting alpha, beta and gamma radiations. 5.1. Auxiliary and counting equipment A certain radiation level is always present because of the "background" produced by cosmic radiation, the radon in the air and the

presence of minute amounts of radioactivity in structural materials and in the earth. This background level of radiation establishes a lower limit for accurate counting if the counting rate of the sample is small. The background level changes from day to day, depending on weather conditions (for example fresh rain dissolves and removes the radon in the atmosphere and thereby lowers the background). To provide a low, more uniform background level for accurate counting a Lead or iron shield is used around the sample and detection tube. Figure 5.1 shows a typical vertical lead shield. Manual counting of pulses is very tedious and can only be used for low count rates. Most counting is performed by instruments known as counters" or "scalers." The counter contains a vacuum-tube amplification circuit similar to those described previously for the various types of detectors (ion chambers, G-M tubes, scintillation wells, etc.). The pulse is amplified to provide sufficient output to operate a mechanical register. However, the maximum rate a mechanical register will accept is about 30 to 50 counts per second. To permit counting at a higher rate, an electronic instrument called a "scaler" is included in the measuring system. The "binary" scaler permits only one count in every 2, 4, 8, 16, 32, or 64 to be recorded by the mechanical register. The use of a scale in which the register records once for every 64 counts is common. Scalers called "decade" scalers are also much used and are based on factors of 10 rather than factors of 2. In using a scaler the number indicated on the register is multiplied by the scaling factor to give the total number of counts for the period that the scaler is operated. Figure 5.2 shows a typical scaler.

5.3 Figures 5.1 S (yo......................... F~~~~~~~~~~~~~~~~~~~~~-i-?iigr 5.2' S c — 4 ~ iiii~ i~-aler ( C u r t es of::: Aoic-::-~~:-:::::::::_:Insrue n Co.)-::::__:-:::

The scaler shown in Figure 5.2 may be operated on any decade scale of 10, 100, or 1000, or on any binary scale of 16, 641 256, 1, 024, and 4,096, by setting the selector on the face of the scaler. The scaler contains a built-in register shown at the upper left of the instrument. Controls for the voltage regulation, pulse height, attenuation, and automatic or manual operation are also shown on the face of the instrument. An external timer is used with this scaler. More elaborate scalers containing both a built-in timer and register are available for use in counting with G-M and scintillation counters as well as with proportional counters. A great variety of other types of auxiliary and counting equipment is available and is described in the catalogs of the various manufacturers. 5.2. The counting of randomly occurring events (1) The decay of radioactive nuclei occurs spontaneously and seems to be influenced in no way by its external environment. It is impossible to predict the decay of any individual nucleus. It is possible, however, to measure the average number of decays that occur over a given time interval frbm a large population of nuclei and to use these measurements in determining useful information concerning the population. The situation is analogous to the use of mortality tables by life insurance companies. No valid prediction concerning the lifetime of an individual can be made from such a table. However, predictions concerning the average lifetime of a population are quite good and extremely useful. In order that the results of measurements made on the decay of radioactive nuclei be used intelligently, something must be known of the statistics which can be used to interpret the data. Radioactive decay can be described by the relation

N = No 5.1 where 1A is the mean life of a nucleus, No is the number of radioactive nuclei in the source at zero time, and N is the number present at time t. It follows that the average number of decays per second is equal to AN. Consider the case where a certain number of disintegrations from a source of known N are measured over a known period.. The question then arises as to how \ should be computed from such a measurement. Experience indicates that the number of events measured over a fixed time interval fluctuates if the process is a random one; thus, some method is needed for estimating how far a particular measured value of disintegrations departs from the average or expected value. 5.~3. The Poisson distribution Consider a time interval of arbitrary length divided into k equal parts as shown in Figure 5.3. Let the probability of the occurrence of an event in any one interval be the same as its occurrence in any other interval. Call this probability p and make (l/k) small enough so that the probability of finding two events in the same interval is vanishingly small. Now the probability of finding an event in the first interval is just p. The probability of finding an event in the first interval and also an event in the second interval is p2. Similarly, for the first, second, and third intervals one gets p3. The probability of not finding an event in the first interval must therefore be (1 - p). Similarly, for not finding one in the first and the second interval one gets (1 - p)2. Thus, the probability of finding an event in the first n intervals and at the same time not finding one in the next k - n is:

f I/k wide I lW I UNIT Figure 5.3: Time Interval Divided into k Units Eack l/k Wide

5.7 pt pn (1 - p)k-n 5.2 One sees that this is one possible arrangement leading to the finding of exactly n events in the total time interval. There are many other arrangements which also give exactly n events. The number of ways of arranging k objects in groups of n each is: Ck = kS 53 n (k - n)! n' It follows that the probability of finding exactly n events in the arbitrary interval is the product of Eqs. 5.2 and 5.3. Call this probability P(n). Then.(n). k: (p)n (1 - p)k-n 5.4 (k - n): n' Assume that k is very much larger than n. This is legitimate since k can be as large as is necessary, the only requirement being that the product kp is constant. In this case k' t kn(k -n) kn (k - n)i: (k -.)'" 5.5a Hence, () (kp)n (L - p)k-n 5.5b P(n) Consider the term (l - p)k-n. Note that this can be written as (L p)k-n= [( p)k-n/-kp] -kp 5.6 As k becomes very large and p becomes correspondingly small (since kp is constant), it is seen that the expression on the right-hand side of Eq. 5.6 approaches ~-kp. Hence, for k very large and p very small, but the product kp still finite, the probability of finding n events, P(n), becomes

5.8 n _9 P(n) 5.7 Here V has been substituted for the constant kp. The expression on the right of Eq. 5-7 is called Poisson's distribution and is the probability of finding exactly n events in the interval concerned. This distribution is suitable for use in describing radioactive decay where, for example, the number of intervals k of the above might correspond to the number of radioactive nuclei in a source, and p might correspond to the probability that any one of them decays in, say, one second. For a gram of U-238, for example, k is around 1021, but p is very small, about 10-18. However, the restrictions of the above derivation, that is that kp be finite, are still readily met. Note that the product kp corresponds to the AN of Eq. 5.1. It should be pointed out that for large values of both n and 4, and for values of n not very different from Ad, the Poisson distribution is closely approximated by the Gaussian distribution. This may be written as G(n) l1/(2rro)l/2 -(n V)2 5.8 In contrast to the Poisson distribution, which describes integral numbers of events, the Gaussian distribution describes a continuous variable. It is, however, a good approximation for the Poisson distribution for large n and is frequently so used. 5.4. Mean value Suppose one asks for the mean value of n in the interval discussed. This is found by taking each value of n from O to X, multiplying by the probability of that value's occurring, and summing. Formally this is expressed as

5.9 co n = nP(n) 5-9 n-O For a Poisson distribution this is 00 n L non C-5) n=O n' Since the first term of the above series is zero, n= e-5,00 - n-l00 n-l=0 (n - n=O no Or, n = P, 5.10 since E in ~~. n=O no Hence, ' or kp has the significance of a mean value. P(n) is thus the probability that n events will be found in an interval if the rean value of events in the interval is V. Experimentally U is the limit of the number found by looking a large number of times in the interval, summing the numbers of events found, and dividing by the number of trials. In the case of the counting of nuclear radiations, \) might be set equal to NT where Ct is some time interval and N is the average number of counts per unit time. (Note that N here is equivalent to the term AN in Eq. 5.1.) Hence, the probability of counting n events in an interval T when the average number to be expected in the interval is NZ is P(n) (N/nr 5.11 An interesting case is the probability of zero. ( (No) 5.12 P(o) =

This is the probability of zero counts occurring in an interval, if the average count rate is N counts per unit time. It follows that the probability that counts will occur in the interval is 1 - P(O), i.e., P(not zero)=1 P- = L - E(o) = 5.13 If a series expansion is used, this gives _e-N' = N+- (NrL2 + (N+)3 + 3 Hence, if Nt >> (Nt)2, which means that Nt must be much less than unity, P(not zero) Nt 5-.5. Coincidence losses and corrections This result has a practical application in radiation counters which have finite resolving times, since it gives the average number of counts to be expected in the dead interval of a counter with resolving tire if the counting rate of a counter with zero resolving time is given by N. Note that the approximation is what one would intuitively expect, but the additional series terms are not given by intuition. As an example, consider a Geiger counter with a resolving time of 200 microseconds and an expected counting rate of 10,000 counts per minute or (10,000/60) per second. The average number of counts occurring in each dead interval is approximately 200 x 10-6 x (104/60) or 1/30 count. In one minute approximately (1/30) x 104 or 333 counts are missed, on the average. Of course, if this is known, this number is simply added to the total counted to get the nuber that would have been counted had the counter had zero resolving time.

5.11 Let N be the number actually counted per unit time, and rT be the number per unit time that should have been counted with zero resolving time. The relationship between the two is given by = N + N(.t) * 5.14 This gives N = N( N () +.. ) I - NV 2 And since NC is usually much less than unity, - N (L + Nt'). 5.15 It should be remarked that this analysis assumes a constant t or counter resolving time. This assumption is not strictly true for Geiger counters, but is a reasonable approximation as long as N( is considerably smaller than unity. 5.6. Deviations from the mean value The voltage output pulses from a counter represent both the nuclear radiations that are being measured or that one wants to measure, together with any other event which may cause a pulse to be produced by the counter. Among the sources of spurious pulses are the naturally radioactive elements that may be incorporated, perhaps as a contaminant, in the counter walls. Others are cosmic rays. In addition, around a laboratory in which radioactive materials are used, additional unwanted sources may be present as accidental contamination or as laboratory sources. The counts which occur as a result of the events in which one is not interested are referred to as "background." They correspond to "noise" in a communication system. If the source that is being measured can be removed from the neighborhood of the counter, measurements

of the background alone can be made. If the background were constant, and equal to the measured count rate, it could then be subtracted from the count observed, which is due to both source and background to give source alone. However, the background pulses themselves will usually occur at random. If they do not, they are due to some systematic source, say electrical line noise, which can be eliminated by suitable precautions*. Since they occur at random, several measurements of background over the same length of time interval may be expected to produce different numbers of counts, but usually of the same order of magnitude. By definition the mean value of a background count is the limit of the average of the counts per interval as the number of intervals considered approaches an infinite number. However, since it is impracticable to measure an infinite number of intervals, it is useful to ask for the expected accuracy of a single measurement or the deviation expected. Such information is useful not only as far as background is concerned, but also with reference to measured count rate of a source. Again, since a limited amount of time is available, one is interested in the expected deviation of a single measurement from the mean value. Experience has shown that a useful quantity for describing this expected deviation is the "variance" of a variable, or the square root of the variance, the "standard deviation." For a variable, x, the variance C2 is defined as o2= (x - x)2 5.16 Here the bar over a quantity indicates a mean value. In words, d2 is the mean value of the square of the difference between x and the mean value of x. Hence, on the average, one might expect a deviation equal to the square root of c2.

5.13 Consideration will show that (x _ )2 = x2 _ 2x +; x2;2 517. If the variable of interest can be described by a Poisson distribution, we need only to know the value of x2 in order to compute aO or a2 since x is known from Eq. 5.lQ.0 If x is the variable n, then n2 Z n2P) - n2 -I v n=O n=O n' Since the first term of the series is zero, it may be written as n-l=O (n -1) or or = 2L-2jV 1 (n - 1) Vn-l I nl (n - 1) n-1=O (n - 1) But this is equivalent to n n n=0 n' n=0 '' However, the first term in the bracket is Just V/E- and the second term is C). Thus, n-= v2 +). 5.18 It follows from Eq. 5.18 that: ~2= ') or f= /J7. 5.19

This equation says that for a Poisson distribution the standard deviation is equal to the square root of the mean value. Hence, the larger the mean value, the greater the expected deviation. It is useful, however, to define a percent standard deviation as An =(n) x 100. 5.20 n For a Poisson distribution this is An = x 100 = 100 Hence, as ) gets larger the percent standard deviation gets smaller and, in fact, can be made as small as desired by making - large enough. Since deviations are to be expected in any set of data, the question now arises as to how an average value is determined experimentally. One answer is that as an approximation one may take a set of data and determine an arithmetic average from it, calling this the mean value. Using this experimentally determined mean value as the actual mean value for, say, a Poisson distribution, one can then predict the standard deviation and the percent standard deviation. These are of course dependent on the experimental choice of 0. Another method that can be used is to choose an approximate value of V and for this value plot a Poisson distribution. This is then compared with the experimental distribution. Other values of 9 can be tried until a best fit has been obtained for the data. The value of V that gives the best fit is then the mean value. For other standard methods of determining v and o from a set of experimental data see References 2 and 3.

5.7. The. effect of background on the interpretation of counting data In the counting of nuclear radiations from radioactive nuclei one usually counts the number of background counts and also for the same time interval the number of counts of source plus background. Suppose ns represents source plus background, background is nb, and n represents source alone. Experimentally, one determines n as n = ns nb. 5.21 The question then arises as to the error to be expected or the standard deviation of an n determined as above. This may be expressed in terms of variances as Cn2= C2 (ns - nb. But this is n2 [(n -5 )-(%h7T) 5.21 o-J) n (ns b)2 -2(ns - nb) (ns nb) + (ns n b)2 Expanded, the right-hand side gives ns2 - 2 nsnb +n 2 ns 2 nb 4 nsnb n 2 nsn nb This gives, when consolidated, n2 (n2 _ n2) + (nb2 nb2) which is equivalent to o 2 = a 2 + a2 n s b

and thus, =s2 +2 5.22 n - Note that if source and background conform to Poisson distributions, Cln >S + nb An assumption frequently made for computing standard deviations approximately is that n =n 8 s and rb ':lb That is, the measured number of counts is almost equal to the mean value for source and background. Hence, ns is expressed as ns + -ns and nb as nb + nb. It follows, according to Eq. 5.22, that if one uses this system, the count is expressed as n = n - n +n+n 5.22 The percent deviation to be expected is thus, n =ns + nb x lOO/(nS - nb) 5.23 The count rate may frequently be of more interest than total counts. Let rate be denoted by caps. Then N. +ns + nb= n + Nt t t t- t t The percent deviation expected in the rate determination is AN = (N/t + Nb/t) x 00 /(Ns -Nb). 5.24

5.17 Implicit in the above discussion is the assumption that the background is counted for the same period of time as is the source. This may not be true. Consideration will be given to the case in which the counting times are different. Note that In2= (n - Ntn)2 5.25 is an alternative definition of the variance of n, where tn is the time interval over which N is computed. However, n = ns(tn/ts) - nb(tn/tb~) Substituting into Eq. 5.25 gives n= CLns (tn/ts) - nb(tn/tb) - Ntn] 2 Substitution of (Ns - Nb) for N and appropriate rearrangement give =2 = (tn/ts)2 [ns N t52] + (tn/tb)2 [I N tb2]0 which is equivalent to n2 = (tn/ts)2 os2 + (tn/tb)2 c2. 5.26 Note that when tn = ts tb this is the same as Eq. 5.22. Assuming that the variables follow a Poisson distribution and further that as2 = Nsts and a%2 = Nbtb, Eq. 5.26 becomes: 1n2 = tn2 (Ns/ts + Nb/tb). 5.27 This expression shows the effect of counting times on the expected error of the result. Note that large counting times produce small values of standard deviation per unit time. Several investigators have studied the

most efficient use of counting time for the achievement of minimum error. For detailed results the literature on the subject should be consulted (4). 5.8. The standard deviation The "standard" deviation for a Poisson distribution of events is equal to the square root of the mean value. The probable error is defined as 0o.67a. In the case of different sets of events, the resultant standard deviation is represented as oN and is given by: i2 ( t ~~~5.28 2 Z N2 n2 5.28 where: aN = Resultant standard deviation. = Standard deviation for the set of events having n mean value in a certain time interval. N = Resultant of the mean values of different sets in different time interval. Assume nl and n2 to be the numbers of the counts taken in time intervals t1 and t2 for a disintegration event represented by a Poisson distribution. Resultant of mean values N = nL + n2 The square of the resultant 2 K)(n1 + n2) 2 9(nl + n2) 21 or the standard deviation an { n On1 +n2 n2 Assuming independence of nL and n2 =N2- nL2 + an22 5.29 oN2= J i / ---

5.19 Magee (10,12) has developed a method of using the reciprocals of the counts so as to limit the effects of geometry on an uranium ore assay. Consider the case in which a counter is located on both sides of a conveyor belt carrying a source. Pulses are recorded from each counter alternately. Let the pulses in a fixed time of integration be equal to nl and n2. Resultant N = i = L-+ n + n2 2, nln2 ~2 n2nln + nn2 2 O-N-~)nL n2+ L ) X nd (n j + n2n n n-n2 2 nln24 + n 4n2 o~ (nl + n2)4 5.30 Case 1. nL =2 = n 2 h n oN =; N = - N 8 2 N 1 N 2n 53 Case 2. If nL >> n2 N = n2 as - is small nL oN2 = n2 N 2

5.20 Thus, the statistical fluctuations are dependent upon the lower count. In many cases, the output of the pulses is recorded by a rate meter through a resistance-capacitance (R-C) network. For such a case let N be the mean count rate and q be the charge per pulse fed to condenser in the time interval between t and t + dt. Charge fed to condenser = qN dt The standard deviation = of this charge The electric charge decays exponentially in a R-C network with a time constant equal to RC. The standard deviation at time tI = q iTdt e-(tlt)/RC 5.33 The total standard deviation on the output at time tL is: 2 = q2N2(tl-t) = I q2Ne 1-t) -00 2 q2 NRC q q- 2 '' 9q = q 5.34 The relative standard deviation 2 Q total charge Total charge = qNRC

5.21 0 - /F2 5.35 Thus, in the case of the use of a rate meter for recording the output, the standard deviation depends on the time constant of the electric network feeding the meter in addition to the mean count rate. Figure 5.4 shows the standard deviation curve and Figure 5.5 shows schematically the effect of the number of counts on the probability of error. 5.9. Percentage probable error The percentage of probable error is given in the case of N-total individual counts as: % Probable Error = (67.45) x (Relative Standard Deviation) % P.E. = 67.45 oN N * P.E. = 6745 5.36 In Figure 5.6 the percent probable error (% P.E.) is plotted versus the total number of counts. In the case of the use of a rate-meter counting device, the percentage probable error is given as: P.E. 67.45 5.37,2 'RC 5.10. Corrections in beta counting The ultimate purpose of counting is to measure disintegration rates. In the counting process, for various reasons, counting rate is generally not equal to the disintegration rate. That is:

5.22 THE STANDARD ERROR IN PARTICLE COUNTING 0 N-/i' N N+J.i MEASURED COUNTS (a) TNE PROBABILITY CURVI... IF UNKNOWN N IS MEASURED AS a, 2/3 OF REPEATS EXPECTED TO BE WITHIN n~/_+ USAEC-ID 20oA Figure 5.4: The Standard Error in Particle Counting NEED FOR LARGE COUNTS TO REDUCE E1101 100% 75 % 4,' _/ /_STAhUAD am(8%) - im1, GBo 10 100 1,000,o 000 MEASURED COUNTS HIGH PROBABILITY OF LOW ERROR REQUIRES HIGH COUNT Figure 5.5: Dependence of Error on Counts

5.23 Counts per min = E (disintegrations per min) where: E ~ unity. The factor E is sometimes referred to as "geometry" but in this text will be referred to as "overall efficiency." Several physical factors listed in Table 5.1 affect the inequality of counting rate and disintegration rate. Some of the factors listed in Table 5.1 are shown schematically in Figure 5.7. Table 5.1 Factors Influencing Counting dead time ge ometry tube efficiency process efficiency scattering absorption sample thickness particle energy 5.11. Dead time After a counting system receives and records a count, there is a finite time during which the system is insensitive to further counts. This is called dead time and is shown schematically in Figure 5.8. When a count is lost owing to the arrival of a particle at the sensitive volume of the counter during dead time, this is called a coincidence loss. Sensitive volume is defined as that volume in which a passing particle may cause a count. There are two experimental means in common use for measuring coincidence losses. In one method, a sample whose half-life is known with great precision is counted as a function of time. The theoretical curve (dotted in Figure 5.9) is extrapolated from the lower counting rate portion of the experimental curve as shown (5).

5.24 NUMBER OF TOTAL COUNTS Figure 5.6: Percent Probable Error vs. Total Number of Counts CONSIDERATIONS RADIOACTIVITY' MEASUREMENTS.WINOW AUSOO11101 AIra AmsornoI BiAC SCATTIERISN ~""~ scsSILE AuOTIORN ARE & MAK ISTAI, -USAEMC-ID 20A Figre 5.7?: Considerations in Radioactivity Measurements

5.25 DEAD TIME OF G-M COUNTER 2 PULSE Is PULSE! + +++;:t~ii iJ;I I::- ----— H i TIME.if -- RECOVERY TIME DEAD TIME 1000 SEC. HEAVY POSITIVE IONS SLOY TO CLEAR. REDUCE EFFECTIVENESS OF FIELD NEAR WIRE. REDUCE SPEED OF ELECTRONS. REDUCE SIZE OF NEXT AVALANCHE Figure 5.8: Deat Time in a G-M Counter 10000 z \ o10 TIME, t - Figure 5.9: Schematic Diagram of Coincidence Losses Due to Dead Time

5.26 Referring to Figure 5.9, the difference between the two curves ((f) is the coincidence loss. For usual dead times, of the order of 200 microseconds per count, the coincidence correction is negligible below 4000 counts per minute. The second method proceeds as follows. If N particles per second pass through the sensitive volume with the production of enough ionization to cause a count, and n particles per second do cause counts, then N-n counts per second are lost. If T (dead time) is time lost per count, then nT is time lost per second and NnT is counts lost per second (5,6). Therefore, N-n = NnT, from which N = n 1 - nT Using two sources, each alone and together, and substituting, n n2 n12 1 - n1 T 1 - n2T 1 - n2 T Eliminating N1 and N2 and solving for T, n1 + n2 - n12 T = - 2 5.38 2 n 2 1 2 T is a function of applied voltage, age of tube, operating temperature and electronic circuit elements. 5.12. Geometry Geometry is defined as that fraction of the particles from a source which start out in the direction of the sensitive volume of the counter (7).

5.27 Referring to Figure 5.10, consider a point source (S) being counted through a circular opening (W). The computation is simple, the geometry factor is merely the ratio of the area of the cap intercepted by the opening to the area of the sphere. A Area of cap k / 2w r RdE but r = R sin G 0 A 7.. k = j 2~ R2 sin d 2 R2 (1 - cos A) 0o Area of sphere K = 4V tR2 g = k/K = - cos A 5.39 where g = "geometry" When the size of the source is not negligible compared with the other dimensions as indicated in Figure 5.11, the function Just derived is subjected to an area integration over the area of the source (7,8). This integration involves an infinite series and yields (7): G =1 1 /.3. __.. 2 [ (L +,)l/2 8 (1 +/- )5/2 Y 6 (L + f)7/2 64 (+/" )9/2 -r- - 5. 040 where G = "geometry" with area integration

C) q3 crd Cb a,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~C 0 0 0 ~~~CI~~~ cc 0 cO CLI -.p (3.) Or)' ~~0 0)~~~~~~~~~~~~~~~~~~~~~~~

5.29 b2 c2 where: b= 2 and 2 = The series converges if r is not too large. If )is allowed to approach zero, G reduces (7) to: ~ 2 ( +6 )LG/ ( - cosA)g 5.41 Accurate experimental determination of the influence of geometry depends on the previous determination of several of the other factors listed, which will be discussed. 5.13. Tube efficiency There exists a finite probability that a particle of the type being measured will pass through the sensitive volume of the counter without causing a count. The probability that a count will be caused by such a particle is defined as tube efficiency (5,6). In accord with present knowledge regarding the discharge mechanism of Geiger tubes, it may be assumed that the formation of one ion pair within the sensitive volume of the tube will produce a discharge. If x is the average number of electrons produced by a particle as it passes through the sensitive volume, the probability that such a particle can pass through producing no electrons is e-x. The probability that a count will be caused by such a particle (tube efficiency) is, therefore, 1 - e-x. If specific ionization is known for the particle to be detected, x can be easily evaluated. The specific ionization s is defined as the number of ion pairs per centimeter of path per atmosphere pressure formed by the passing particle. The total average number of electrons so formed then is sLp, where L is the average path length through the counter and p is pressure of the gas in atmospheres. From which (5):

5.30 Eo = 1 - e LP 5.42 Consider the case of an argon-filled counter at 0.1 atmosphere pressure. Then s = 30, and if L = 2 cm, E= - e30(2)(0.1) = - e-6 = 99.8% In general, for accurate work, an Eo of less than 98 percent is not tolerable. For a given energy particle, sG varies linearly with number of electrons per atom of the gas. Thus, to increase efficiency, gases which yield a higher specific ionization are used, and pressure, or the average path length is increased. The simplest way to do the latter is to move the source farther away from the tube. A straightforward means of measuring Eo involves a triple coincidence circuit. Since efficiency is defined as the ratio of number of observed counts per unit time to the number of ionizing particles passing through the sensitive volume of the tube in that time (e.g., E = No/n)t one can state that, for a triple coincidence circuit, Eo =E E E2 E3 =Nt/n 5.43 E1, E2, E3 are the respective efficiencies of the individual tubes and Nt is the observed counting rate. If the center tube is then disconnected and left in place, E= E1 E3 = Nd/n, 5.44 where Nd is the observed counting rate. From Eqs. 5.43 and 5.44: E2 = N/Nd 5.45

5.14. Process efficiency Inefficiencies in sample preparation and in the counting process must be known and reproducible for accurate absolute counting. Some typical inefficiencies are: (i) adsorption on glassware during sample preparation, (ii) loss of volatile components during evaporation, (iii) spattering during evaporation, (iv) and lateral shifting of the source during a series of counts. In the case of counting in which the aim is to measure relative counting rates, these inefficiencies need not be considered if both the unknown and the standard are prepared and counted in the same manner. Losses due to adsorption may often be kept to a negligible amount by techniques such as acidification of the solution, by repeating rinsings of glassware after each operation, and by the addition of carrier (nonradioactive isotope of the same element). The loss of volatile components is inevitable in evaporation processes, but it may be estimated by chemical or radiological survey of the escaping gas and often may be minimized by the addition of suitable chemicals. S~pattering is best kept at a minimum by very slow evaporation and may be estimated by a radiological survey of the area surrounding the evaporating specimen. Lateral shifting from the center has the effect of reducing the angle subtended at each point of the source by the window. The magnitude of the effect is best determined experimentally for a particular counting arrangement. Figure 5.12 shows some typical results of displacement from the center (7). When an absorption curve is taken with the absorbers near the source, a maximum will appear at the early portion of the curve. The effect increases with the atomic number of the absorbers and also increases as the absorber is brought nearer the source. This is called forescatter (8).

5.32 Absorber just above 12 source (4th she) I8 z Absorber 2nd shelf I I 4 8 12 16 THICKNESS OF POLYSTYRENE, mg/cm2 Figure 5.13: Percent Forescattering As Determined with Polystyrene PER CENT OF MAXIMUM COUNT.0 (, I I I O0 -4 -1 -I Figure 5.12: Effect of Asymmetry on Efficiency of Counting

5.33 Polystyrene, which has about the same atomic number as air, can be used to make an estimate of forescattering due to the air between source and window. In this procedure successive thicknesses of polystyrene film are inserted at various positions between source and window. The percent of counting increase over the count with no polystyrene is plotted against film thickness in Figure 5.13, The initial slope-of these curves is a measure of percentage increase per unit thickness and gives the differential scattering effect for each position. An integration of this differential scattering effect between source and window gives an estimate of the forescatter factor due to the air thickness. This factor increases with source-to-window distance and decreases with energy, and varies for usual counting conditions from 1.001 to 1.060. Obviously the effect is best minimized by having absorbers as close as possible to the tube, ~5 16. Backscatter The counting rate has been found to depend on the backing upon which a source rests. A measure of the effect may be obtained by comparison with the counting rate of a source backed by a film of the order of 50 micrograms/cm2, for which the backscatter factor is effectively unity. As polystyrene film was added, the factor was found to go iip to 1.01 -.1.05. The factor increases with increasing energy and with increasing atomic number of backing material. Figure 5.14 shows the variation of backscattered radiation as a function of atomic number and energy (8). Figure 5.15 illustrates how the backscatter coefficient approaches its maximum value with the addition of backing material (8). The maximum value iof backscatter coefficient, as approached asymptotically in Figure 5.15, is referred to as saturation backscatter. It

5.34 z 80 I<~~~~~~~~~ ~~p 32 60 RaE Uo 40 I w 20 20 40 60 80 ATOMIC NUMBER, Z Figure 5.14: Backscatter as a Function of Atomic Number and Energy of Radiation 50 U. o 0 0 20 40 60 80 THICKNESS OF POLYSTYRENE 6, mg/cm2 Figure 5.15: Saturation Backscatte

/ i \ 5.35 is characteristic for a specific energy and backing material. The value of the coefficient varies from 1,1 for paper to 1.8 for platinum. An additional contribution to backscatter is presented by the source support structure and the inner walls and floor of the lead shield. To measure this effect, a tube and a source support structure of negligible mass were arranged and the various components of a typical counting mechanism brought to their normal positions. The integrated effect upon counting rate was taken to be the contribution to backscatter due to housing alone. It was determined experimentally that this factor differs appreciably from unity only on shelves near the floor of the housing. In this case, the factor had values as high as 1.10. 5.17. Se if-scatter When the source is thick and the radiation from a point within is considered, it is obvious that both fore- and backscatter by the source, itself are involved. The magnitudes of the two effects can be estimated., if the average atomic number is known, from the work of the two previous sections. In actual experimental practice, however, these are extremely difficult to measure separately. Their effects are often lumped into measurements of self-absorption. 5.18. Self-absorption As the source thickness is increased, absorption within the source material itself assumes importance. In general, the self-absorption factor would merely be the ratio of counting rate for the thick sample to counting rate for a source of zero thickness. There are two experimental approaches to this measurement. In one case, one prepares sources of varying thicknesses containing the same amount of radioactive material. The resulting curves are similar to the usual absorption curve, as

5.36 indicated in Figure 5.16. The slight downward concavity at low sample thickness is the same phenomenon exhibited at low absorbers of forescatter. The self-absorption coefficient, of course, can be read directly from the ordinate for the particular element being studied at the sample thickness involved. In the other case, samples of equal specific activity but varying sample weights are prepared and counted. For samples of zero thickness, the self-absorption coefficient, of course, is unity. This is a very desirable state of affairs, but is unattainable in general practice. Another simple situation occurs when the source is thick compared with the range of the beta particles involved. Such a sample is called "infinitely thick." In this case, specific activity is proportional to the radiation emitted per square centimeter of surface. An advantage of using samples of infinite thickness is that, for a particular particle energy and absorber, specific activity is directly measurable. Also, the larger mass of sample material is more precisely measurable to exacting degrees of precision. It is advisable for such measurements to have a large quantity of radioactive material, since only a small fraction of the betaparticles originating in the source can be detected. The quantities are, however, not always available. In the region of absorber thickness between zero and infinite thickness (approximately): dAm = ex/P dAt where dAt = -cldx (The contribution of each additional increment of thickness). x At = e dje- x/ dx =(Q4 (1 - e-x/B) A l - e-x/a At x/,u 5.46

5-37 where Am = measured disintegration rate At = actual disintegration rate { = specific activity x = thickness of sample )a = absorption coefficient (mg/cm2) T = surface area of sample. This relation, of course, must be consistent both as x approaches zero and as x exceeds particle range. In the former case: at x = 0 A = 0, At = 0 5.47 Since the specific activity is constant, the count must go to zero with decreasing thickness. In the latter case: Lim Am 5.48 x-+u At x;, R where R = Irange:. Making x greater than R obviously will have no further effect upon Am/At. Increasing the energy will tend both to increase particle range and decrease the absorption coefficient, which gives the expected reduction in the self-absorption coefficient. Increasing atomic number of the absorber, on the other hand, will tend to increase the absorption coefficient and thus increase the self-absorption. 5.14. Window and air absorption The magnitude of the factor for absorption in the air and in the tube window is evaluated by extrapolation of an aluminum absorption curve to zero absorber, as shown in Figure 5.17. The ratio of counts under experimental conditions to counts for zero absorber is the factor for air and window absorption. It is essential

5.38 I 4 0 0 -J 4 0 W0 ___0 4 LO (n 0 OO 50~~~SMPLE THICKNESS mg cvlcm Fjgure 516: Se~f~Pb8orptiOn 0. C') I110I20 30 TOTAL ABSORBER THICKNESS, mg/Cma F'igure 5.17: Asorpt)iofl jP ir and. Win&ow

5.39 that very high total counts be used in the low absorption region of the curve in order to avoid extrapolation errors. This linear extrapolation is sound since, for most beta particles, absorber thicknesses of the order of 6 to 10 mg/cm2 are on the linear portion of the curve. This, of course assumes that forescatter has been minimized. For a rigorous measurement of low-energy particles, such as from carbon-14, air and window absorption must be considerably reduced. 5.20. Energy The dependence of overall efficiency on energy is not direct, but shows itself rather in the energy dependence of the various factors previously discussed. Experimentally, however, overall efficiency is known to vary with energy. At energies greater than 1 Mev this variation is slight, but at lower energies the overall efficiency falls sharply. This is primarily due to the increasing effect of window and air absorption. As low energies are approached, there will finally appear a critical energy, below which no counts can be recorded. This "threshold" energy will vary with window and air thicknesses. Of the foregoing, the following are demonstrably energy-dependent: tube efficiency -- increasing energy means greater velocities and therefore less specific ionization and reduced efficiency. forescattering -- increasing energy increases the forward momentum of the particles and thereby reduces the effect. backscattering -- increasing energy enables more of the backscattered particles to enter the counter and increases the apparent effect. self-scattering -- combination of two previous effects. absorption -- increasing energy reduces the probability of interaction and reduces absorption effects.

5.40 In the case of assaying techniques, it is generally possible to make a precise determination without a detailed knowledge of the foregoing factors. In most such techniques, energy and sample weight must be independent variables. In order to determine overall efficiency as a function of energy and sample weight, however, it is possible to hold other factors constant. If standards of the same energy as the unknown are prepared in the same way on the same backing and counted in the same counter for roughly equal total counts, then the only corrections which need be made are those due to energy and sample thickness or weight variation as shown in Figures 5.18 and 5.19, respectively (9). 5.21. Relative importance of correction factors in beta counting The geometry factor is always important since it describes the maximum percentage of the total number of particles which can be counted. It varies from approximately 50 percent for a windowless flow counter to values ranging from about 30 percent to 1 percent or less for end window counters. The absorption factor may vary for a thin window (1.8 mg/cm2) tube from approximately 1.05 for energetic betas, As from phosphorus-32, to as high as 1.2 or 1.3 for weak betas, such as from carbon-14. With thicker windows or longer air paths, the absorption factor may be much higher for weak beta emitters. Similarly, the effects of self-absorption become appreciable with the weak beta emitters. The information derived from most experiments is the relative disintegration rates or activities between samples. The efficiency can be made quite high, approaching 100 percent for gamma rays with a scintillation detector and suitable arrangements of phosphor and source. The

.AJ t~ I: >0 w "o J w 0 0 I z L J LL. WI-oW!R 4" O 00 ENERGY, Mev Figure 5.18: Effect of Energy on Beta Counting 1.0.~~..~ ~-'"'""~-~. _.,._,. P-32 w 0.95 LL. w RoE 0 0.90 z 0 Co -60 a3 C: 0.85 0 I. U._ w!.8 0 20 40 60 80 I00 120 140 160 SAMPLE WEIGHT, mg. Figure 5.19: Effect. of Sample Weight on SelfAbsorption Coefficient in Beta Counting

5.42 scintillation detector can usually be made much more efficient for the detection of gammas than can the Geiger counter. In each case the geometry will depend on the physical arrangement used. For relative measurements, all the correction factors need not be known if they can be kept constant. This means in all cases using a fixed geometry and mounting all the samples in the same way. In the case of the weak beta emitters, one must be sure that all the samples are of the same thickness (preferably of infinite thickness) and in the same chemical form since the self-absorption factor will depend on the density of the material also. For example, in carbon-14 tracer experiments it is quite common to convert all samples to barium carbonate before counting. 5.22. The counting of gamma emitters G-M counters can be used for gamma emitters, but their efficiency is very low (around 1 percent) compared to their high efficiency for betas. Nevertheless, when the activity levels are high enough, G-M tubes are used. Special tubes have been designed incorporating high density plates made of bismuth metal within the counting volume in order to increase the gamma interaction probability. Such tubes have efficiencies as high as 4 or 5 percent. If it is necessary, or expeditious, to use an ordinary Geiger counting tube as a gamma detector, then the following information may be helpful. If it is assumed that no photons enter the counting volume except through the thin window and that the effect of scattering on counting rate may be neglected (57): c/m = E (d/m) = g (d) (t) (3.7 x 107 x 60) Eo 5.49

5.43 where E = overall efficiency (p. 5.21) g = geometrical factor (eq. 5.39) ~= number of millicuries ~= number of photons per disintegration Eo = tube efficiency (Eq. 5.42). If it is assumed that tube efficiency for gamma counting is linearly related to photon energy, ' (58): Eo - 0.00oo69 5.50 and that the calculated counting rate is conveniently plotted in Figure 5.20 as a function of the ratio of source to counter distance to counter radius (cot A from Figure 5.10). If a counting sample is so thick that self-absorption must be considered, but not thick enough to introduce a significant variation in the geometrical factor "g" throughout the sample thickness, the following simple correction can be made (57): N l - e-PXo = 5.51 N1 JXo where No = counting rate with no absorption N = counting rate with absorption correction = linear absorption coefficient for the sample medium and photo energy and Xo = sample thickness. For greater efficiency a gamma scintillation counter is often used. These counters usually have gamma efficiencies of 20 - 50 percent, The high-efficiency results both from the high densities of some phosphors that are available and from the large possible detecting volumes that can

COUNTING RATE, COUNTS MINUTE 0000 0 O o o o ~ V L t AIXIYA ~~~~~I LlZ z z VI Z Al Z O o SH ~~~~ 0 o @ o,, e en - W ' _. '! i C)~~~~~~~~~~~~[ r.... 1 ', ' I I ' z,z/,I/ ~,, -r- zz, z/h ~ _, i" lYM:",,';,x ~:, 0, 1z3 lo ~~~~~~~~J ~_ -- ".,~ "-.., _:(D.0 - -~ -'. -,'D~~-~-~", -X, I, L_ < I _ '- ' 4~ '" ~ J '~ E.

5.45 be used. The latter point is particularly well illustrated by the plastic and liquid scintillators that have been devised for use in well-type counters in which the source is nearly surrounded by the phosphor. High counting rates with minimum dead time losses are achieved by using phosphors with very short light pulses such as stilbene which has a light pulse decay con-8 stant of about 10 seconds. The resolving time of the counter can thus be made quite low. The most commonly used gamma counter is the well-type scintillation counter. It has the advantage of good gamma efficiency and excellent geometry. The gamma scintillation counter has the additional advantage in that it is capable of energy discrimination, i.e., it can be used in a gamma spectrometer system to distinguish one isotope from another. The sample preparation is much simpler for gamma emitters than for beta emitters since in most cases absorption and self-absorption are negligible. This means that usually there is no problem in regard to sample thickness, chemical form, or density. Absolute gamma counting is usually done by comparison with a known standard of the same isotope or some standard with a long half-life which has a similar gamma spectrum. 5.23. Counting alpha particles The efficiency of scintillation detectors may be made very nearly 100 percent for alphas and for betas with energies of about 100 key or more. However, owing to their very low penetrating power, alpha particles are generally counted in a windowless flow counter, operated either in the G-M or proportional region, or with some form of ionization chamber.

5.46 Alpha emitters are rarely used in tracer experiments. The only elements having alpha-emitting isotopes are polonium, thorium, uranium, radium, and elements of atomic number 206 and higher. Additional information on counting statistics is given in References 10 and 11 and on counting in general in References 13-56.

Chapter 5 References 1. Kerr, W., "Nuclear Engineering Measurement" in Nuclear Reactors and Radiations in Industry, 1, The University of Michigan, Ann Arbor, Mich., 1966. 2. Nichols, H. and Rauch, L. L. "Engineering Measurements and Instrumentation," Automatic Control Course Notes, The University of Michigan, Ann Arbor, Mich., June, 1955. 3. Feller, W., "An Introduction to Probability Theory and Its Applications," John Wiley and Sons, Inc., N.Y., 1951. 4. Loevinger, R., and Berman, M., "The Efficiency Criteria in Radioactivity Counters " Nucleonics, 9, No. 1, 26, 1951. 5. Korff, S. "Electron and Nuclear Counters," D. Van Nostrand, Co., Inc., N.Y., 1946. 6. Hoag, J. B.,and Korff, S., "Electron and Nuclear Physics," D. Van Nostrand, Co., Inc., N.Y., 1928. 7. Burtt, B., "Absolute Beta Counting," Nucleonics, 5, No. 2, 28, 1949. 8. Zumwalt, L. R., "Absolute Beta Counting Using End Window GeigerMller Counters and Experimental Data on Beta-particle Scattering Effects," AECU-567, Oak Ridge National Laboratory, Oak Ridge, Tenn. Sept. 14, 1949. 9. Cowan, F. and Nehemias, J., "Sensitivity of the Evaporation Method of Liquid Waste Monitoring," Nucleonics, 7, No. 5, 39, 1950. 10. Shaw, E. N., "Statistics of Radiation Measurements," Nuclear Engng., l, No. 4, 152, 1956. 11. Fry, T. C., "Probability and Its Engineering Uses," D. Van Nostrand Co., Inc., N.Y., 1928. 12. Magee, K. W., Austronic Engineering Laboratories, Melbourne, Australia, unpublished. 13. Lorenz, E., Weikel, J. and Norten, S. G., "A Counting-Plate and Frequency Meter," Rev. of Sci. Instrum., 17, 276, 1946. 14. Kip, A., Bousquet, A., Evades, R. and Tuttle, W., "The Design and Operation of an Improved Counting-Rate Meter," ibid., 17, 323, 1946. 15. Schultz, H. L., "A Frequency Meter for Random and Uniformly Spaced Pulses," ibid., 18, 223, 1947. 5.47

5.48 16. Elmore, W. C., "Electronics for the Nuclear Physicist - Part III - A Counting-Rate Meter," Nucleonics, 2, No. 4, 43, L945. 17. Kleopper, R. M. and Hoecker, F. E., "A Double-Channel Direct-Reading Low-Frequency Counting-Rate Meter and Counting-Rate Comparitor," Rev. sci. Instrum., 20, 17, 1949. 18. Cooke-Yarborough, E. H., "A New Pulse-Amplitude Discriminator Circuit," J. sci. Instrum., 26, 96, 1949. 19. Rose, M. E., and Korff, S. A., "An Investigation of the Properties of Proportional Counters," Phys. Rev., 59, 850, 1941. 20. Simpson, J. A., Jr., "A Precision Alpha-Proportional Counter," Rev. of sci. Instrum., 18, 884, 1947. 21. Corson, D. R. and Wilson, R. R., "Particle and Quantum Counters," Rev. of sci. Instrum., 19; 222, 1948. 22. Putman, J. L. "Analysis of Spurious Counts in Geiger Counters," Proc. phys. Soc., Lond., 61, 312, 1948. 23. Curran, S. C. and Rae, E. R., "Analysis of the Impulses from GeigerMtfller Tubes," Rev. of sci. Instrum., 18, 871, 1947. 24. Jurney, E. T. and Maierschein, F., "The Gamma-Ray Counting Efficiency of a Lead-Cathode G-M Counter," Rev. sci. Instrum., 20, 932, 1949. 25. Miller, W. W., "High Efficiency Counting of C14 as C02," Science 105, 123, 194r7. 26. Cooke-Yarborough, E. H. and Pulsford, E. W., "A Founting-Rate Meter of High Accuracy," Proc. Instr. elect. Engrs., 98, 191, 1951. 27. Cooke-Yarborough, E. H. and Pulsford, E. W., "An Accurate Logarithmic Counting-Rate Meter covering a Wide Range," ibid., 98, 196, 1951. 28. Rotblat, J., Thomas, D. G. A. and Sayle, E. A., "Scale-of-Hundred Counting Unit," J. sci. Instrum., 25, 33, 1948. 29. Taylor, D. "Count and Time Control in Radiometric Assay," ibid., 81, 1950. 30. Taylor, D., "Radioactivity Surveying and Monitoring Instruments," ibid.., 81, 1950. 31. Taylor, D., "Electronic Instrumentation in Atomic Research," Engineering, Lond., 169, 631 and 644, 1950. 32. Greinacher, H., "Spark Counter for Counting Corpuscles and Photons," Helv. phys. acta., 9, 590, 1936.

5.49 33. Chang, W. Y., and Rosenblum, S., "A Simple Counting System for AlphaRay Spectra and the Energy Distribution of Po Alpha-Particles," Phys. Rev., 67, 222, 1945. 34. Keuffel, J. W., "Parallel-Plate Counters," Rev. sci. Instrum., 20, 202, 1949. 35. Neddermayer, S. H., Althus, E. J., Allison, W. and Schatz, E. R., "The Measurement of Ultra-Short Time Intervals," ibid., L8, 488, 1947. 36. Kallman, H. and Broser, I., "Die Erregung von Phophoren durch schnelle Teilchen," Z. Natur., 2a, 439, and 642, 1947. 37. Jurney, E. T. and Maienschein, F., "The Gamma-ray Counting Efficiency for a Lead Cathode Counter," Rev. sci. Instrum., 20, 942, 1949. 38, Novey, T. B., "Ra DEF Standards of Absolute Activity Measurements," UAC-104, AECU-947', U.S. Atomic Energy Commission, Wash., D.C., May 31, 1949. 39. Pannell, J. H., "Radioactivity Measurement Techniques," Massachusetts Institute of Technology Document, AECD-2270, U.S. Atomic Energy Commission, Wash., D.C., Nov., 1947. 40. Reiss, M., Badrick, F. E., Halkerston, J. M. and White, J. H., "A Method for Continuous Graphic Recording of Radioactive Tracer Concentrations from Various Body Regions Simultaneously," Biochem. J., 44, 255, 1949. 41. Sinclair, W. K. and Newberry, S. P., "A Direct Reading Meter for the Measurement of Highly Active Samples of Gamma-emitting Radioisotope, J. sci, Instrum., 28, 234, 1951. 42. Taylor, D., "The Measurement of Radioisotopes," Methuen and Co., Ltd., Lond., 1951. 43. Brownell, G. L. and Lockhart, H. S., "CO2 Counter Techniques for C-14 Measurement," Tech. Report No. 30, Lab. for Nucl. Sci. and Engng., Mass. Inst. of Tech., Cambridge, Mass., 1949. 44. Bruceb, M., King, E. R., and Bruner, H. D., "A Method for Standardization of Gallium 72," ORO-44, U.S. Atomic Energy Commission, Wash., D.C., 1951. 45. Burch, G., Reaser, P., Ray, T. and Threefoot, S., "A Method of Preparing Biologic Fluids for Counting Radioelements," J. Lab. clin. Med., 35, 626, 1950. 46. Burch, G., Reaser, P., Threefoot, S. and Ray, T., "A Micropipette for Preparation of Samples for Counting in Radiobiology," J. Lab. clin. Med.., 35, 631, 1950. 47. Cannon, C. V., "Conference on Absolute Beta Counting, Preliminary Report No. 8," Nuclear Science Series, National Research Council, U.S. Gov. Print. Office, Wash., D.C., 1950.

48. Feiterlberg, S., "Standardization of Radioactive Iodine," Science, 109, 456, 1949. 49. Freedman, A. J. and Hume, D. N., "A Precision Method of Counting Radioactive Liquid Samples," Science, 112, 461, 1950. 50. Goodwin, W. E. and Harris, W. D., "A Method for the Determination of Small Doses of I-131 in the Urine," J. Lab. clin. Med., 38, 470, 1951~. 51. Hill, R. F., Hine, G. J. and Marinelli, L. D., "Quantitative Determination of Gamma Radiation in Biological Research," Amer. J. Roentgenol., 63, 160, 1950. 52. Jordan, W. H., "Detection of Nuclear Particles," Ann. Rev. Nuclear Sci., 1, 207-244, 1952. 53. Keene, J. P., "An Absolute Method for Measuring the Activity of Radioactive Isotopes," Nature, Lond., 166, 601, 1950. 54. Kirby, H. W., "Determination of Tracers in the Presence of their Radioactive Daughters," Analyt. Chem. 24, 1678, 1952. 55. Loevinger, R. and Berman, M., "Efficiency Criteria in Radioactivity Counting," Nucleonics, 9, No. 1, 26, 1951. 56. Mayncord, W. V. and Roberts, J. E., "An Attempt at Precision Measurements of Gamma Rays," Brit, J. Radiol., 10, 365, 1937. 57. Aiba, S., "Effects of Various Geometrical Factors on Gamma-Ray Counting," J. sci. Res. Inst., Tokyo, 49 144, 1955. 58. Sinclair, W. K., "Comparison of Geiger-Counter and Ion-Chamber Methods of Measuring Gamma Radiation," NucLeonics, 7, No. 6, 21, 1950.

Experimental Techniques in Nuclear Tracer Studies By A. Gordus 6.1 Introduction (By L. E. Brownell) In the field of research no other tool since the invention of the microscope has been as useful as radioactive "tracers" in extending knowledge. Many important results have been obtained during the few years that radioactive tracers have been used and many more will be obtained in the years to come. During the past twenty years, and particularly since the last great war, thousands of reports have been published on the use of tracer amounts of radioisotopes for diagnosis in medicine, for control of variables in industrial processes, and for research studies in science. Most of these reports are technical and describe particular applications in specific and widely different fields. It is difficult, therefore, to comprehend the great importance of the new tool of radioactive tracers. Tracer methods of analysis and research would appear to be "made to order" for studies in the biological fields. The use of tracer techniques has resulted in remarkable progress in the fields of medicine and agriculture. Photosynthesis is one of the most important but has been one of the most baffling processes in the latter field. A better understanding of this process has been possible as a result of the use of tracer methodso The role of mineral nutrients in the soil has been extensively studied by tracer techniques. It has long been known that plants require appreciable quantities of phosphorus, nitrogen and potassium, together with trace amounts of other elements for normal growth. In the past the requirements for soil fertilization have been judged largely from crop yields. However, other variables such as rainfall, temperature and disease complicate such procedures. By tracer methods the capacity of the soil to supply a plant, nutrient can be determined by adding a labeled fertilizer to the soil as a partial source of the nutrient. Tracer analysis has been employed in a wide range of fields for solving research problemso In all the branches of pure and applied science, tagged atoms are being used to obtain quantitative and qualitative information. Studies of chemical reactions, rates of diffusion of metals, vulcanization of rubber, and water analysis are some selected examples of fields in which the radioactive isotopes have been used to advantageo Some of the most useful radioisotopes in such research are carbon-14, sulphur-35, and tritium, an.isotope of hydrogen. 6ol

Tracer analysis has been used for wear studies in the tool and the machinery industries,for the study of mass transfer in the transistor field, for the detection of minute amounts of impurities or residues, for oil prospecting in the petroleum industry, for the determination of the optimum time of mixing in the chemical industry, and for a multitude of other industrial applicationso The principle of analysis using radioactive tracers has received much publicity but the story of Go Hevesy, the originator of the method, may not be as familiar to the average scientific investigator, even though Hevesy received the Nobel award in 1943 for this important contribution to science. In 1911 Hevesy was working with Rutherford in England where he.attempted to separate-RaD from lead by chemical meanso At that time, it was not known that these two materials were isotopes and therefore could not be separated by chemical means but only by physical means. As a result of these studies Hevesy correctly concluded that RaD participated in every chemical reaction in which lead was involved. In 1913 at the Vienna Institute of Radium, Hevesy used this information to measure the solubility of nearly insoluble lead salts, such as the chromate and sulfide, using lead having some natural radioactivity. As a result of his studies in the use of radioisotopes in chemical analysis, Hevesy in 1932 proposed the technique of analysis by isotope dilution. Two years later the Joliots reported artificial radioactivity in elements exposed to alpha-particle radiation from natural radioisotopes This was followed by the development of particle accelerators such as the cyclotron and then the nuclear reactor, which made available new radioisotopes covering the entire range of atomic numbers. This opened the door to the use of the tracer method which became a new tool in the study and analysis of many different systems. With the availability of a means of producing.artificial radioactivity, Hevesy in 1936 also pioneered the development of another tracer technique known as activation analysis. Seaborg and Livingood in 1938 reported studies on the technique of activation analysis. The method of isotope dilution is still the widely used tracer technique, however. Other methods such as reverse isotope dilution, derivative dilution, activation analysis, and autoradiography have greatly increased the range of application of radioactive tracers. Most of these methods are discussed by Hevesy in his book.

6.2 Outline for first three weeks of study (A.A. Gordus) The first three weeks of the course will be devoted to the discussion and measurement of radiations of various types, i.e., gamma, beta, and alpha. The following experiments will:be performed in 'the laboratory section during this time. First Week:. Determination of Plateau. Coincidence Counting'References: (* - Optional) Overman and Clark, p. 25-31; 51-61; 84-86*, 259-262; 271-274* Glasstone, p. 136-140 (p's 6.25-6.39); 291 (P 11.24); 477 P'sl 17.9-17.10 Friedlander and Kennedy, coincidence correction 265-266 determination of 273-274* coincidence counting 244-245 for disintegration rate measurements '95-296* in decay scheme studies 284* coincidence resolving time 245 coincidence spectrometry 246*, 284* in study of Aul98 decay 286-287* coincidence, delayed 141-142, 245 Geiger-Mueller counters, 231-235 computed efficiency of 2.64-265 dead time of 234 discharge mechanism of 233-234 filling mixtures for, 234 plateau of 233 portable.248 self-quenching 234 6.3

-2 -Second Week:' Absorption of Radiation References: Overman and Clark, (see previous page) Friedlander and Kennedy (op. cit ) absorbers, placement of 207-208*; 296* useful types of 273 absorption coefficient, for beta particles 4*, 198 for gamma ray-s. 206 absorption corrections in' absolute -counting 292, 294-295, 296* absorption curves effect of backscattering on 276 extrapolation to zero absorber 296* for beta particles 197-203 for conversion electrons 201-203 for gamma rays 208 for (x)-rays 208; 210-211 absorption edges for (x)-rays 205, 209-211 overall combination of.above pages 141-142; 197-203; 205-211; 231-235; 244-245; 246*; 248; 264-265; 265-266*; 273-274*; 276; 284*; 286-287*; 292; 294-295; 296*. Half-life, comparative 167-170* definition of 7, 127, 236 determination of 8, 127-129, 140-143. effect of chemical state on.166 from decay curves 128-129, 140-141 from decay'-systematics 142-143 from delayed coincidence 141-142 from specific radioactivity 142 importance in decay scheme studies 283 partial 136-137 for alpha decay 175-176 for beta decay 169-170 for gamma emission 155-156 for spontaneous fission 180, 181 6.4

-3-.-Glass tone, S.., Half-life, biological, defined 508*:(P 18.29) radioactive, defined 120-122 (P 5.46-.... 49) determination 122-123 (P's 5.50 -5.55) and minimum dectectable quantity 1 447 (l's 16.34 16.35) Glasstone, S. p. Cit.) p. 149-151 (i's: 7.1-7.7); 151-153 (O's 7.8-7.13);. 65- 171 (Pi's 7.67 -7.78) overall combination of above pages for first two weeks: p. 136-140; 149-153; 163-171; 291; 477. Third Week: Absorption of Radiation (Continued) Measurement of Half-life References: (Continued) Half-life Qverman and Clark, radioactive, of components of mixtures 290-300 definition of 287 determination of, from -coincidence measurements 1309-310* by decay.of artificial nuclides, Expt. 8-1 323-324* by decay of naturally occurring nuclides, Expt. 8-2 324-326* from decay measurements 290-295, 298-300, 304, 311, 314-316, 3'23-337* by differential - count method 299 -300 of genetically related nuclides, secular equilibrium, Expt. 8-6 333-337* transient equilibrium, Expt. 8-5 330-333* by growth measurements, Expt. 8-7.337-338* from growth measurements 307-308, 311-313, 315-316, 330-338* of independent components of mixture, Expt. 8-4 329-330* by method of averages 292 6.

-4 -by method of least squares 292-293 from reaction yield -322 for short,.lived nuclidet Expt. 8-3;26-329* by specific. activity measurements, Expt. 8-8 3'38-342* from specific - activity measurements 296-297 estimatio'n from radiation energy 297-298 of (n,y) produced radionruclides - table 4.55 —459 partial 289-290 reliability of 293-295 6.6

6.3 GENERAL LABORATORY RULES FOR NUCLEAR CHEMISTRY The following rules must be observed in the group to insure efficient and contamination-free laboratories and equipment. 1. All work with radioisotopes is to be done in the hood and not on the bench. top. Ratioisotopes are not to be removed from the hood except in the form of mounted samples, stoppered tubes or packaged sources. 2. All samples taken into the counting room must be "tied down" to prevent spread of contamination; solids must be covered with plastic, scotch tape, etc.; tubes of liquid must be stoppered. The outside of these samples must be absolutely free from contamination. 3. For all measurements with the well counter the sample tube must be placed inside a second plastic tube to protect the crystal from contamination. Similarly- extreme care must be used with the scintillation spectrometers. (Crystals cost a lot of money and take a long while to replace!) 4. Extreme care must be taken to guard against contamination of electronic equipment in the counting rooms. Check your hands for contamination before operating the equipment. Be sure prepared standards and samples are properly mounted and safely stored between runs. 5. Radioactive wastes should be accumulated in cartons and jugs inside the hood and then transferred to the waste storage can or cupboard in the laboratory. The Radiation Control Service (RCS) at extension 2-592 will pick these wastes up. Radioactive wastes should include any wastes that are possibly contaminated. 6. Gloves are to be used in 'the hood whenever the hands would contact radioactive materials. Gloves or other possibly contaminated objects are not to be removed from the hot lab. 7. Procedure for cleaning contaminated equipment should include: rinses in the hood, soak in active cleaning solution, and finally wash in an active sink. Materials used to work with high (millicurie) levels of activity must be kept separate from other equipment and so labelled. 8. Glassware and other materials are not to be put back in the general stock drawers or cupboards unless the user is positive that they contain no contamination. When in doubt place with possibly contaminated items. 9. Stock solutions of carriers in plastic bottles are not to be used at the bench. They should serve merely as reservoirs to fill individual dropping bottles with a few ml for use in the hood.,..~ ~ 6.

6.4 SCALE OF 64 0 1.2 3 4 5 6 7 8 9 0 00000 00064 00128 00192 00256 00320 00384 00448 00512 00576 1 00640 00704 00768 00832 00896 00960 01024 88 01152 01216 2 01280 01344 01408 01472 01536 01600 01664 01728 01792 01856 3 01920 01984 02048 02112 02176 02240 02304 02368 02432 02496 4 02560 02624 02688 02752 02816 02880 02944 03008 03072 03136.5 03200 03264 03328 03392 03456 03520 03584 03648 03712 03776 6 03840 03904 03968 04032 04096 04160 04224 04288 04352 04416 7 04480 04544 04608 04672 04736 04800 04864 04928 04992 05056 8 05120 05184 05248 05312 05376 05440 05504 05568 05632 05696 9 05760 05824. 05888 05952 06016 06080 06144 06208 06272 06336 10 06400 06464 06528 06592 06656 06720 06784 06848 06912 06976 11 07040 07104 07168 07232 07296 07360 07424 07488 07552 07616 12 07680 07744 07808 07872 07936 08000 08064 08128 08192 08256 13 08320 08384 08448 08512 08576 08640 08704 08768 08832 08896 14 08960 09024 09088 09152 09216 09280 09344 09408 09472 09536 15 09600 09664 09728 09792 09856 09920 09984 10048 10112 10176 16 10240 10304; 10368 10432 10496 10560 10624 10688 10752 10816 17 10880 10944 11008 11072 11136 11200 11264 11328 11392 11456 18 11520 11584 11648 11712 11776 11840 11904 11968 12032 12096 19 12160 12224 12288 12352 12416 12480 12544 12608 12672 12736 20 12800 12864 12928 12992 13056 13120 13184 13248 13312 13376 21 13440 13504 13568 13632 13696 13760 13824 13888 13952 14016 22 14080 14144 14208 14272 14336 14400 14464 14528 14592 14656 23 14720 14784 14848 14912 14976 15040 15104 15168 15232 15296 24 15360 15424 15488 15552 15616 15680 15744 15808 15872 15936 25 16000 16064 16128 16192 16256 16320 16384 16448 16512 16576 26 16640 16704 16768 16832 16896 16960 17024 17088 17152 17216 27 17280 17344 17.408 17472 17536 17600 17664 17728 17792 17856 28 17920 17984 18048 18112 18176 18240 18304 18368 18432 18496 29 18560 18624 18688' 18752 18816 18880 18944 19008 19072 19136 30 19200 19264 19328 19392 19456 19520 19584 19648 19712 19776 31 19840 19904 19968 20032 20096 20160 20224 20288 20352 20416 32 20480 20544 20608 20672 20736 20G$00 20864 20928 20992 21056 33 21120 21184 21248 21312 21376 21440 21504 21568 21632 2-1696 34 21760 21824 21888 21952 22016 22080 22144 22208 22272 22336 35 22400 22464' 22528 22592 22656 22720 22784 22848 22912 22976 36 23040 23104 23168 23232 23296 23360 23424 23488 23552 23616 37 23680 23744 4 23808 23872 23936 24000 24064- 24128 24192 24256 38 24320 24384 24448 24512 24576 24640 24704 24768 24832 24896 39 24960 25024 25088 25152 25216 25280 2525344 25408 25472 25536 40 25600 25664 25728 25792 25856 25920 25984 26048 26112 26176 41 26240 26304' 26368 26432 26496 26560 26624 26688 26752 26816 42 26880 26944 27008 27072 27136 27200 27264 27328 27392 27456 43 27520 27584 27648 27712 27776 27840 27904 27968 28032 28096 44 28160- 28224 28288 28352 28416 28480 28544 28608 28672 28736 45 28800 28864 28928 28992 29056 29120 29184 29248 29312 29376 46 29440 29504 29568 29632 29696 29760 29824 29888 29952 30016 47 30080 30144 30208 30272 30336 30400 30464 30528 30592 30656 48 30720 30784 30848 30912 30976 31040 31104 31168 31232 31296 49 31360 31424 31488 31552 31616 31680 31744 31808 31872 31936 6.8

0 1 2 15 4 5 6 7 8 9 50.32000 32064 32128 32192 32256 32320 32384 32448 32512 32576 51 32640 132704 32768 32832 32896 32960 33024 33088 33152 33216 52 33280 33344 33408 33472 33536 33600 33664 33728 33792 33856 53 33920 33984 34048 34112 34176 34240 34304 34368 34432 34496 54 34560 34624 34688 34752 34816 34880 34944 35Q008 35072 35136 55 35200 355264 35328 355392 35456 35520 315584 35648 35712 35776 56 35840 35904 35968 56032 56096 56160 56224 56288 56352 56416 57 36480 36544 56608 6672 56736 56800 56864 56928 56992 57056 58 37120 57184 57248 57312 57376 57440 57504 57568 57632 57696 59 37760 57824 57888 57952 38016 78080 78144 78208 78272 78336 60 38400 38464 38528 38592 38656 38720 38784 38848 38912 38976 61 390401 39104 39168 392352 39296,39360 39424 39488 39332 39616 62 396801 397441 39808 39872 39936 40000 40064 40128 40192 40256 63 40320 40384 40448 40512 40576 40640 40704 -0768 40832 408F9 64 40960 41024 4o1088 41152 41216 41280 411344 41408 41472 41536 65 41600 41664 41728 41792 41856 41920 41984 42048 42112 42176 66 42240 425304 42368 42432 42496 42560 42624 42688 42752 42816 67 42880 42944 43008 43072 43136 43200 43264 43328 43392 43456 68 43520 43584 43648 43712 43776 43840 43904 43968 44032 44096 69 44160 44224 44288 44352 44416 44480 44544 44608 44672 44736 70 44800 44864 44928 44992 45056 45120 45184 45248 45312 45376 71 45440 45504 45568 45632 45696 45760 45824 45888 45952 46016 72 46080 46144 46208 46272 46336 46400 46464 46528 46592 46656 73 46720 46784 46848 46912 46976 47040 47104 47168 47832 47296 74 47360 47424 47488 47552 47616 47680 47744 47808 47872 47936 75 48000 48064 48128 48192 48256 48320 48384 48448 48512 48576 76 48640 48704 48768 48832 48896 48960 49024 49088 49152 49216 77 49280 49344 49408 49472 49536 49600 49664 49728 49792 49856 78 49920 49984 50048 50112 50176 50240 50304 50368 50432 50496 79 50560 50624 506686 50752 50816 50680 5094T 5100 51072 51136 80 51200 51264 513128 51392 51456 51520 51584 51648 51712 51776 81 51840 51904 51968 52032 52096 52160 52224 52288 52352 52416 82 52480 52544 52608 52672 527536 52800 52864 52928 52992 53056 83 53120 53184 53248 53312 53378 53440 53504 53568 53632 53696 84 53760 53824 53888 53952 54016 54080 54144 54208 54272 54336 85 54400 54464 54528 54592 54656 54720 54784 54848 54912 54976 86.55040 55104 55168 55232 55296 55360 55424 55488 55552 55616 87 55680 55744 55808 55872 559356 56000 56064 56128 56192 56256 88 56320 56384 56448 56512 56576 56640 56704 56768 56832 56896 89 56960 57024 57088 57152 57216 57280 57344 57408 57472 57536 90 57600 57664 57728 57792 57856 57920 57984 848 58112 58176 91 58240 58304 58368 584312 58496 58560 58624 58688 58752 58816 92 58880 58944 59008 59072 59136 59200 59264 59328 59392 59456 93 59520 59584 59648 59712 59776 59840 59904 59968 600352 60096 94 60160 60224 60288 60352 60416 60480 60544 60608 60672 60736 95 60800 60864 60928 60992 61056 61120 61184 61248 61312 61376 96 6i440 61504 61568 61632 61696 61760 61824 61888 61952 62016 97 62080 62144 62208 62272 621336 62400 62464 62528 62592 62656 98 62720 62784 62848 62912 62976 63040 631504 63168 63232 63296 99 653360 635424 63488 63552 635616 635680 63744 63808 63872 63936 6 9

6.5 Experiment 1. Counter Operation (Adapted from G. Wilkinson) A,. Plateau and operating voltage: In turning on a scaling unit, the master switch is first turned on, with the high voltage knob turned to the "Off" position. Scalers are usually left on for some time before use to obtain thermal equilibrium and stable operation. It is of utmost importance that the scaler be allowed. to warm up at least one minute before the high voltage -is turned on. If this procedure is not followed, the Geiger tube may be'damaged by going into discharge. To start operating, the high voltage switch is turned on. The count switch should be on and a radioactive source in the counter. When the high voltage has warmed up, as evidenced by reading of the meter, the voltage is raised slowly until the counter begins counount. This voltage is the "starting voltage ", or "threshold", which has a value of about 600-1200 volts. If the counter is counting at more than 15,000 counts per minute (240 scales of 64 or 120 of 128 per minute)' the source should be lowered. Geiger tubes are likely to count erratically at rates that are too high. In order to select a good "operating" voltage at which to count, it is necessary to obtain a "plateau". This is done by taking a series of counts at intervals of 25-50 volts from starting voltage up for 250 or 300 volts and back down again. If the counts are plotted vs. voltage, a curve is obtained which should have a relatively flat portion which extends down to within 50 or 100 volts of starting voltage. The operating voltage should be in the plateau region near its lower end. Usually this means between 75 to 150 volts above' the starting voltage. The operating voltage for a proportional counter is found in a similar way, but there is no sharp starting voltage. NOTES: (1) In the chlorine quenched "Amperex" tubes used in these experiments an excessive voltage does no appreciable damage. However, with more conventional counter tubes which are filled with argon plus an organic quenching substance such as alcohol, an excessive voltage will quickly ruin the counting tubes. For such tubes, if the scaler does not begin to register counts by the time the voltage on the tube 6. 10

has been raised to, say, 1500 volts, check over the connections and ensure that the rest of the apparatus is working. Plot a graph for the plateau determination as you proceed: you should find that a sharply rising portion is followed by a more or less level 'plateau'. After this, the counting rate will again rise fairly-steeply. This region is to be avoided since here the counter tube is fairly rapidly damaged, due to the excessive rate of decomposition of the polyatomic component of the filling gas. In any case this region is useless for accurate work, because of the incipient instability which it represents. Counts in excess of 15,000 per minute should be avoided. An alcohol quenched counter will not'last indefinitely. It should operate satisfactorily for a total of about 108 counts. Various chan'ges are likely to occur. The starting voltage usually increases, the background rate increases and the plateau becomes less flat. A plateau which has a slope of 5. or less per 100 volts is good. If the slope is more than 7X, it is bad. A tube that has deteriorated is likely also to display erratic counting as discussed below. (2) Counter tubes. Some of those provided are of the end-window type, having a thin window of mica sealed to the lower end to permit the entry of reasonably weak beta particles. Please note: 1. The windows are extremely fragile, and should not be handled at all. The tubes cost $355.00 - $50.00 each and are very easily ruined. 2. As normally run, the outer case (cathode) is at ground potential, and the wire (anode) at about 1000-1500 v. positive with respect to ground. A piece of fine copper wire is used for the anode connection: make sure this does not touch any grounded metal. Turn off voltage before handling counter tube since the high voltage could give a very unpleasant shock. 6..II

-3 -35. At all costs avoid the possibility of radioactivity contaminating the inside of the lead housing, or, even worse, the sets of aluminum absorbing screens. Scrupulous care is called for in preparing and handling sources. 4. Counters should not be left unnecessarily with a source in place, or with the potential on the wire, since this shortens their useful life. It may be desirable at times, however, to leave the potential on the wire to prevent "warmm-up" errors. At no time, though, should a sample be left in place when the counting is finished. 5. Remember that the stop-count switch merely connects the scaling stages to the tube so that the discharges can be counted. The tube still;'counts" a sample when the count switch is off. 6. If a countler appears not to work at its normal voltage, be cautious in increasing the applied potential, and first check that the fault does not lie in some other part of the circuit of the tube. B. Counting Techniques The simple counting of radioactive samples requires a calibration of the counter consisting of (a) taking a plateau and selecting the operating voltage, (b) measuring the dead time. Each sample measurement involves the following steps: (a) Measurement of the background rate. (b) Measurement of the sample-plus-background or "total rate". (c) Correction of the total rate for dead time. (d) Subtraction to find the rate due to the sample alone, the "net rate". (e) Estimation of the probable error in the determination of the net. rate. (f) Correction for day to day change in counter efficiency. The procedure for sample measurement is then as follows: (a) Suppose that XB background counts, say 605, are registered in a time tB, say 30 minutes. The background rate NB is 6.12

-4 -XB given by the fraction, N = B (b) Suppose also that X total counts, say 5261, are registered in a time ts, say 2 minutes. The total rate is Ns = Xs/ts, in this case 5261/2 = 2630.5 c/m. (c) The correction for dead time is read from the table which has already been made up for the counter. For the rate of 2630 c/m, it may be 48.2 c/m. This is added to give the corrected total rate, N' = 2630 5 + 48.2 2678.7 c/m. c-o ~re 3.0~ 5 ~ 7 c/ (d) The net rate M is N's NB or 2678.7 - 20.2 = 2658.5 c/m. (e) For rates of ten times background or more, the standard deviation in per cent is given to a sufficient degree of accuracy by the formula 100i Xs This formula can be plotted as p.er cent against total counts to facilitate the determination of the standard deviation. Estimation of errors for slower counting rates is somewhat more complicated and is discussed later. it must not be overlooked that Geiger-Miller tubes count at an appreciable rate in the absence of radioactive sources. This rate due to cosmic radiation and natural contamination is called the background, and may be around 50-60 c/m in a clean laboratory. Heavy Pb shields cut the background to 15-25 c/m, and make the tubes nearly insensitive to activity being measured on nearby counters. (f) The efficiency of counters often changes with time. Thus counts of samples must be no:rmalized by counting a reference standard long-lived source regularly. After both sample and reference counts are corrected for background and resolving time giving M and R c/m, the ratio M/R gives the normalized count. Usually one reference count per 15 minutes is adequateif variations outside the limits of error do not occur, the reference counts can be averaged over a period of say a day, before using the value for normalizing. Make a complete measurement of several samples furnisheld you. Discuss the decay scheme and its relation.to the counting rate of each isotope measured. 6013

Experimentla. Determination of Counting'Rate Losses by Extrapolation of a Decay Curve. (Adapted from G. Wilkinson) The dead time: After a Geiger counter has counted a particle, there is a short period of time (of the order of 1/3 millisecond) during which.no particle can be counted. 'For this reason, at high rates the count is less than the number of particles entering the sensitive region of the counter. The dead time of the counter may be measured and -the total counting rates' may be corrected for the loss. The correction varies as the square of the counting rate. The reason for this.is' that the:number of counts: lost is proportional to the product of the number of dead.interv.als and the number. of particles which will arrive during' a -single interval. Each of these factors increases with the counting rate. As an example, if 8 c/m must be added to a rate of 1000 c/m, then 132 c/m must be added to 2000 c/m and 200 c/m must be added to 5000-c/m. A special graph or table is usually made up for correcting counts from each counter when its dead time has been measured.. This is a plot of the measured counting rate (abscissa) counts per minute vs. the correction for -dead time in counts per minute'. Background rates are low enough that they require no dead time correction. Proportional counters have practically no dead time and so counts made with them usually require no dead time correction. The coincidence loss or dead time can be determined in several ways. One method outlined below involves the direct determination of counting rate losses by extrapolation of a known decay curve. The second main method is that known as the "split source-" method. Procedure A source of radioactive material is prepared so that it has a counting rate of about: 30,000 c/min. Place in-'a suitable counter shelf or, if a solution, in a solution jacket, and start 'counting. Take a succession of counts at one minute intervals; the. best procedure is probably to count for 50 seconds, from each minute, leaving'ten seconds for noting the scaler reading. Leave the. stopwatch running throughout this experiment, and then plot observed activity against time on a semi-log paper. Draw a straight line fitting it to the lower points on the graph. Plot also the discrepancies at the higher rates of counting, -..!4

-2 -i.e., extrapolated value minus the observed value, expressed as a percentage of the true (extrapolated) counting rate, as a function of the observed counting rate, and deduce the counter "dead time". Determine the half-life of the material. If the observed counting rate is R, and the resolving time w, then the true rate R' is given by R' = R (R counts are observed in 1 - RT second, since RT out of each second is 'dead time') R so Rr= l - Rr - R.., = R i.e. slope of f discrepancy vs. R curve.= 6t15

Half-life - determination of The data obtained in the'above section lends itself to the problem of half-life determination. The data for dead time correction was made using a hot sample of an unknown element. The data as collected and plotted above yield directly the half-life of the' unknown radioisotope. What is the half-life? Can you determine the isotope involved? What is the equation of its formation and decay assuming bombardment with thermal neutrons from the nuclear reactor?

6.6 Experiment 2. Backscattering See Special Dangers listed at end before beginning work!! Nucleons are emitted from radioactive substances in a random manner and only a very -small portion of these disintegrations are observed unless special tubes are used which completely surround the sample. The particular spatial arrangement used in measuring a sample with a tube will influence the efficiency of the counting. Also, the material used to hold the sample is 'very influencial since radiation is easily scattered by matter resulting in an increase in the number of particles actually passing into the counting tube and being observed. This scattering must be controlled if meaningful results are desired. Scattering has been observed to be a function of spatial arrangement, atomic number of the backing (scattering) material backing thickness and the energy of the radiation. To measure these affects, begin by noting the background. Then place a radioactive beta emitter on a shelf such 'that the counting rate is about 4000 counts per minute. (This number is chosen so that effect of backscattering should be easily observed.) Make all countings for five minutes then take average. Place any absorbers used below the sample so that there is always a constant distance between the sample, backingS, upper surface, and tube. One may have to tane additional plates under the first absorber when increasing backing thickness. See Figure 1. Since scattering is also dependent on the backing's atomic number, try two or more different materials like plastic, copper, lead, aluminum, etc. Plot a curve of net counts per minute minus background vs. (Z), the atomic number, of scatter for saturated condition (over 250 mg per cm2 thick), 6 t7 ~

ccounting tube sampi e constant distances sample holder (card with hole in center) first scatterer with others added below it.plastic shelving Figure 1 Determine the atomic number of an unknown material by utilizing backscattering. Can you suggest an analytical application for backscattering? Plot also the counting rate vs. the thickness of a backing material. Discuss any ramifications of these facts as they would apply to designing equipment and experimental procedures in radio-tracer analysis. As time permits. Determine what relationship exists between beta energy and backscatterer by using other beta sources if they are avail.able and measuring the counting rate increase for any infinitely thick scatterer with various energy beta particles. Table I 1 source tZ particals emitted energy T1204 4.1y - 0.765 Pm'47 2.52y 5- 0.223 p32 24.4d A- 0.248 I13z 8105d1 - < ' a...... 131 8 (od'-'a-s-) 0 5 '(0.608 87% CS 37 0 0.335 9.3% Cs)'..0..y. o f.*..{0.,523 92% V1.19.8% Special Dangers 1. Be careful not to contaminate absorbers. (You should record the background before and after all experiments to determine if you have been careless and contaminated the equipment. You will be required to decontaminate your errors!) 2. Be careful not to break or scratch tube window. The windows are thin and puncture destroying the expensive tube very easily. Use extra special care when working on top shelf. 6,18

6.7 Experiment 3 (Adapted from G. Wilkinson) A. Radiochemical Separation of Bi210 from Pb210 Lead separated from uranium ores contains a very small amount of the radioactive isotope Pb210 (T! = 22.2y). This decays through the 2 chain: 82Pb10 10 L 82Pb (stale) The principle activit measu22.2re 5d ith a Geiger er counter is The principle activity measured with a GeigerMller counter is that due to the n-rays from Bi210 (Ema 1.17 Mev) since the very soft n-rayys from b20 and the s from Po20 are readily absorbed. Gamma Radiation from this chain is negligible. The specific activity of the lead is low, however, and it is desirable to isolate the Bi210 for counting. The stock solution of lead contains 0.10 g. of Pb(N03)2 and 0.0002 g. of Bi carrier (as the nitrate) per ml. of 0.15 N HN03 solution. Pipette 1.0 ml. of the stock solution into a 50 ml. centrifuge tube. Add dropwise a solution of NaaC03 until the precipitate which forms dissolves with difficulty. This operation neutralizes the solution. Dilute the solution with 25 ml. of H20 and add 0.5 g. KBrO3. Boil the solution and add dropwise 1 M HNO3 to clear the solution of any cloudy precipitate. To the hot solution add 50% KBr dropwise until the solution is deep brown. Boil until the Br2 is removed. A small precipitate of BiOBr should be present. Repeat the KBr addition and boiling until no more precipitate forms and the solution is a clear yellow. Three additions of KBr should be adequate. Centrifuge for about 2 minutes while still hot and decant the supernatant solution. Wash twice with 5-10 ml. H20 centrifuging for about two minutes each time. Drain thoroughly by inverting the centrifuge tube on a paper towel for about 5 minutes. Dissolve the precipitate in 0.1 ml. of 1 M HC1, and transfer to source mount with a capillary pipette (drawn from 6-8 mm tubing). Evaporate on special hot plate. Rinse the centrifuge tube with another 0.1 ml. of 1 M HC1 and add to the source and evaporate to dryness..19

Prepare a second source with 0.10 ml. of the Pb stock solution. Cover both sources with polystyrene films before counting. Source mounts are prepared by fastening: 1mil polystyrene film over the hole in pre-cut counting cards. To do this put a thin layer of rubber cement over one face of the card and allow to dry throughly (2-3 minutes). Press a- 2" square of the polystyrene film on to the card so that it is held taut over the 1" hole. The samples are than counted on the 2nd shelf of the counting unit. Three sources will be used: 1. your separated Bi20 source 2. a stronger Bi210 source, which will be available in the count-ing room. 3. your Pb210 -Bi210 source B. Determination of the-Al Absorption Curve of Bi210 (RaE) The radiations from a source may be qualitatively identified from the Al absorption curve, and the 5 energy of prominent constituents identified. Itt will be found that the absorption curves vary somewhat with the geometrical disposition of source and Al sheet absorbers with respect to the counter window. Accepted techniques involve placement of the source on the second or third shelf and the absorbers on the shelf immediately below the counter window. Determine the counting rate with Al absorbers of roughly the following weights in mg. per cm2: 0, 7, 20, 40, 70, 100, 150, 200, 300, 500, 700, 1000. The counting rate of the source will be reduced over 100-fold' for the last few absorbers, and it is not feasible to count such a wide range of rates with one source. When the counting rate has fallen below 400 c/m replace your source No. 1 with source No. 2, after carefully determining the ratio of the counting rates of both sources through some convenient absorbers. The portion of the curve obtained With the second source can then be plotted on the same curve with your original source. Plot your net activity (c/m) as ordinate against weight of absorber (mg/cm2) as abscissa on 3 cycle semi-log paper. Be sure to 6 020

-3 -include in the absorber the weight of polystyrene cover film ( 3 mg.), the weight of the air between sample and counter and the weight of the counter window. Extrapolate your curve to zero on the abscissa to obtain the counting rate of the sample with zero absorber. Extrapolate the curve to l"zero" counting rate to obtain the visual range of Bi210 D rays (476 mg/cm2). From your extrapolated range in Al calculate Emax for Bi210 5 rays. Use this curve to construct a Feather Analyzer (see L. E. Glendenin, p. 17, Nucleonics, 2, January 1948). Measure sources No. 1 and No. 3 under the same geometrical conditions through approximately 3.0 mg. of absorber and from these two measurements calculate the efficiency of your chemical separation. A slight error will be introduced by the thickness of the Pb(N03)2 source. How will this error affect your yield calculated? If one assumes that the overall counting efficiency for a sample on the second shelf is 4%, calculate specific activity (d/m/mg Pb) of the lead.in the stock solution. Repeat the measurement of aluminum absorption with another radioactive isotope which will be provided. Determine the visual range, and by the use of a Feather analysis the more accurate range. From range-energy relation curves determine the maximum energy of the D particle measured. Notes on Aluminum Absorption Measurements (1) Any penetrating radiation which is present (nuclear gamma rays, or secondary electromagnetic radiation -- 'Bremsstrahlung') must be allowed for by extrapolating back to the intensity axis any hard component, and subtracting the extrapolated values from the observed total intensity at each point. This procedure is exactly analogous to that for resolving a complex decay curve, by subtraction of the longer lived component, obtained by extrapolation, from the total activity at any time.. o21

-4 -The curve should eventually turn downwards, and go more or less asymptotically to a value of the thickness of aluminum which represents the range of the beta particles. In the case of pure betaemitters, this end-point will usually be fairly clear, but in the presence of appreciable gamma ray backgrounds, or with weak sources, the uncertainty in the measurements at lower intensities may mask this effect entirely, and a procedure due to Feather must be adopted, in which the more accurate, early part of the absorption curve is compared with that for a 'standard' substance (usually RaE). (2) The relationship between energy of beta particles and their range is a fairly. definite one, although frequently difficult to use on account of uncertainties in the estimation of the range. It is important to 'standardize' each individual counting set up with beta emitters of known energy, before reliance can be placed on any results obtained. (3) The maximum energy of the n-particle spectrum of a radioactive substance is a valuable parameter for identifying the isotope responsible for the emission. It is only rarely that equipment (magnetic lens spectrograph, for example) for making an absolute measurement of this quantity is available in an ordinary laboratory, and one must therefore have recourse to indirect, comparative methods. A good measure of the D-particle energy can be obtained by measuring the range of the particles, or, alternatively, their absorption coefficient (or equivalent half-thickness) in some suitable material. That normally chosen is aluminum, since this is about the lightest material readily obtainable in the form of thin foils. Heavier materials result in the production of greater amounts of continuous X-radiation ('Bremsstrahlung'), which remain as a penetrating background when all the n-particles have been absorbed. To carry out the measurement, increasing thicknesses of aluminum foil are placed in front of the counter window and the counting rate for a constant source is determined. If the source decays significantly during the operation, correction must, of course, be applied. It is most desirable to use a thin source, mounted on a thin backing, and placed so as to minimize reflection of particles into the counter; 6.22

-5 -in this way one can preserve aas s possible the original energy spectrum. The logarithm of the net v-particle count (after deducting any background due to cosmic rays, y-rays, or Bremsstrahlung) is plotted against the thickness of aluminum adding an allowance for the thickness of the counter window, and for the stopping power of the air gap between source and counter; 1 cm. of air at N.T.P. is equivalent to 1.29 mg./cm2.). From the initial, almost linear, part of this curve, the halfthickness or absorption coefficient may be deduced. For a pure 5 -emitter, with no y-ray background present, the range of the Bparticles, which,is a more significant measure of their maximum energy than the absorption half-thickness, can be obtained by inspection. In the presence of a large y-ray background, the statistical accuracy of this part of the curve is inevitably poor, and a method due to Feather (Proc. Ca-mb, Phil. Soc., )4, 599. 1938) is frequently used. In this method a substance of known s-particle range is adopted as a reference substance (RaE, with range 0.476 g/cm2 aluminum is frequently used)- Logarithmic plots of the net a count for both RaE and the unknown are then made, and normalized so that both start from the same point on the log; plot (on semi-log paper this involves only the moving of the unknown curve bodily up or down until it coincides with the RaE curve for zero thickness of absorber.) The known range of RaE Ps is then divided into, say, 10 equal steps, and the ordinates which, for RaE, correspond to each of these abscissae are extended to cut the absorption curve for the unknown substance. The thickness of aluminum, XnRx, at which the ordinate corresponding to n/10 of the RaE range cuts the unknown curve, gives, when multiplied by 10/n, an estimate of the range of the unknown betas. If the two absorption curves were identical in shape, this would be an exact relationship and Rx would be equal to (10/n)xnRX for all values of (n). In practice this is not necessarily true, but this quantity will become a better approximation to the true range as (n) is increased towards 10. In the limit, for n = 10, the two are 6.23

-6 -identical, by definition, but this region is often experimentally inaccessible, and one has to rely upon extrapolation from smaller values of (n). One- can, in fact, plot the apparent range as a function of (n), obtaining a curve which frequently has the appearance shown here; there is little difficulty in then obtaining quite a reliable value of the range of the betas, without having to carry the actual measurements down into the region where the gamma ray background is too large to permit accurate beta particle measurements. With some modification,. this method may also be used for thick sources, but for this the original paper should be consulted. In general, the Feather method gives results which are more or less independent of the counting geometry; this will be true also for ranges obtained by visual inspection of absorption:curves.:In the case of absorption half-thicknesses, however, this is not necessarily true, and it should be borne in mind that the value which one obtains experimentally is, to some extent, dependent on the relative positions of source, absorbers, and detector; and on the nature of the detector (it is clearly somewhat dependent on the way in which the response of the detector varies with n-particle energy). For this reason, quoted values of half-thicknesses should be regarded in a slightly critical light, and not -taken to apply, without further checking, to any other particular measuring system. C. Analysis of y:ray absorption data The purpose of this experiment is to make a preliminary study of the problem of identifying the gamma radiation from typical radioactive sources by absorption techniques and to make semiquantitative studies of the energy and yield of principal gamma constituents. There will be provided in the laboratory stock solution of the following radioactive systems at ~54. c/ml: 5-13 5.3y Co60 8.Od I"'1 with 0.05 mg. carrier/mi. About 1 mg. of Ag+ carrier should be added to the I when it is taken to dryness to prevent volatilization. 6.24

-7 -12.8d Ba140 in equilibrium with 40 h La' o Each student will select one of these systems for study and will obtain the experimental absorption data on each of the others from a colleague. The report will consist of the 5 and y absorption curves of each of the three radioactive systems, plus the graphical analysis, and discussion. Reported values for the yields and energies of the main 5 and y constituents of these radionuclides (Seaborg and Per lman, Table of Isotopes, Rev. Mod. Phys. 20. 585-667 (1948) are: 5.3y Co60 simple n-ray specturm of 0.31 Mev followed by two yrays in cascade of: 1.16 and 1.32 Mev. 8.0d OId 3 85% of 0.600 Mev followed by y of 0..367 Mev; 15% P of.315 Mev followed by a of 0.638 Mev. 12.8d Bal4~ 75% 6 Of 1.0 Mev and no a, 25% of P of 0.5 Mev. followed by y of 0.53 Mev.; 40.0Oh Lal40 (disintegration rate in transient equilibrium is 1.15 times that of Bal40 parent) 70% of.- 1.4 Mev., 20% = 0.90 Mev., 10% = 2.12 Mev. and 77% of y = 1.65 Mev., 12% 0.85 Mev., 6% 0.49 Mev, and 5% 2.3 Mev. 1. P Counting and the ':y.Ratio A small known aliquot of solution (to give ~200 c/s on the second shelf) will be evaporated on a polystyrene film. The first portion of the absorption curve will be taken (o, 10, 20, 40 mg/cm2 of added absorber), together with the activity through an Al absorber greater than the range of the hardest 5 component. This P absorption curve should be analyzed qualitatively to ascertain that it is in agreement with the data reported for the radioactive system. With the aid of this analysis, extrapolate the activity to zero total absorber. Compare this activity to that just beyond the range. This comparison is called the P - ratio. It is a function of the number and energies of the rays. Assuming the Y counting efficiency is approximately linear with v-energy and has a value of 0.5% at i.0 Mev. compute the ':7 ratio for each of the radioactive systems studied, using the decay schemes 6.25

-8 -reported above, and neglecting attenuation of?'-ray in the Al absorber used to reach the 5 range. 2. ~counting and lead absorption curve. Mount an active sample of the stock activity so that when placed in the bottom shelf of the counter with an aluminum absorber of sufficient thickness to stop all @_ rays immediately above the sample, the sample counts about 5-8000 c/m. Try the effect of placing a thin lead absorber (a) directly above the aluminum (b) on the middle shelf (c) as close to the counter as possible. Explain the variation in counting rates observed. Keep the sample in the bottom shelf with its aluminum cover; place a second aluminum absorber immediately below the counter window. This second absorber must be thick enough to stop any Compton or photoelectrons generated by the Y rays; calculate this thickness for the maximum? energy involved. Now place lead absorbers between the aluminum sandwich, plot the activity vs g/cm2 lead absorber, and esti-mate the energy of the y' rays from the half thickness-energy curves for ) radiation. Note: In estimation of? ray energies by absorption methods very serious errors can occur due to scattering phenomena; the best geometry is to have (1) an unshielded counter well away from scattering materials (2) the active source far away from the counter and its beam will collimated (3) the absorbers fairly close to the counter. 6.26

6.8 Experiment No. 4 Radiochemical Separations - Parent-Daughter Decay Separation of Ce'44 from Pr'44 by Precipitation References: Overman and Clark, Chapter 8 and Experiment 8-6, p. 333. Writeup by A. S. Newton on "precipitates". An inactive run should be made on this separation to check the problems involved in the chemical procedure before the active Ce-Pr is used. The longest lived cerium isotope resulting from the fission of uranium is cerium-144. The decay is as follows: Cel44 P-(0.35 Mev) Pr'44 r-(.1 Mev) Ndl44(stable) 285 days 17.3 minutes A small amount of gamma radiation of 0.15, 0.22, and 1.25 Mev is associated with the Pr'44 decay The purpose of this experiment is to chemically separatej cerium from praseodymium activity, and, on the basis of the buildup and decay characteristics of this parent-daughter equilibrium, to calculate the degree of efficiency with which the separation was carried out. The separation is performed in the presence of carriers of Ce and La by the oxidation of the Ce to the plus four oxidation state followed by precipitation as cerium iodate. The precipitation is fairly complete but may carry with it some of the plus three rare earth ions. Pr can be precipitated later as the hydroxide using lanthanum carrier. Because of the fact that the Pr activity grows in very rapidly, it is necessary that as little time as possible pass between the precipitation of the cerium iodate (zero time, in terms of the separation) and the first counting of Ce-Pr "build-up" sample. Procedure Take sufficient of the Ce tracer solution to give between 5000 and 10,000 counts per minute of Pr'44 (i.e. with aluminum absorber, 100 mg/cm2, to shield out the Ce144 activity) on shelf three of the end-window counter. To this solution, in a 50 ml. centrifuge tube, add 10 mg of Ce carrier and 10 mg of La carrier. -Also add 5 ml of conc. HN03 and 2 ml of 0.5 M NaBr03. Allow two minutes for reaction. 6~27

-2 -(Cerium is oxidized rapidly under these conditions, more rapidly and smoothly than with the use of KC103 in boiling HNO3.) Add 20 ml of 0.35 M HI03 (note the precise time-to-the second) stir, and let stand for two minutes. Centrifuge. Transfer the supernatant solution to a spare tube. Wash the precipitate twice with 1 M HN03 plus a little HI03 and add the wash solutions to the supernatant solution. Slurry the precipitate in a small volume of water and mount on filter paper according to the instructions of Newton. Do not bother at this time to dry completely and weigh the sample. Instead, rinse with a few ml of alcohol and evacuate rapidly to dryness. Mount the filter paper in the center of a mounting card and cover with cellophane or mylar film. Interpose an aluminum absorber to remove the CeL44 beta rays. Take as many one-half minute counts as possible until about 20 minutes since "zero time", (the time of separation). Equilibrium between the parent and daughter will occur about one to 1.5 hours after zero time. At this time, take a series of counts to determine the activity of daughter in equilibrium. As soon as the initial "build-up" activity has been determined (i.e. about 20 minutes after zero time) carry out an isolation of the Pr activity removed in the precipitation process. Take a known fraction or all of the supernates plus washings containing the Pr activity. Neutralize this with NH40H, bring to a boil to coagulate La(OH)3-Pr(OH);3 precipitates, centrifuge in a 50 ml tube, and wash with two portions of slightly alkaline solution. Collect the precipitate, evacuate rapidly to dryness, mount, and count. Measure the decay curve of this Pr fraction using the absorber for the Ce44 activity. Conclusions: Calculate the percent Pr144 contamination in the separated Ce as well as the percent Ce144 contamination in the separated Pr. To obtain these data the following calculation should be performed. 1. Utilize the parent-daughter relation: C = (C) extrap(e-At - e-Bt) + CB e-Bt 6,28

-3 -Let us assume that build-up counts of one-half minute duration are recorded. The observed activity, corrected for background and coincidence counting corrections, is multiplied by two to yield a value of the activity in counts per minute. The time, t, relative to the zero time at which any of the one-half minute counts were taken is simply the mid-point of the count. These then are the observed values, CB. (CB)ext:apoIated is the average observed activity of the daughter at equilibrium, since the half-life of the parent is very long relative to the time at which the counts were taken with respect to zero-time. Refer to Figure 1. For each of these build-up counts, a value of CB may be calculated. Due to the fact that the calculation will involve the subtraction of two large values, and the numerical difference will become smaller as the time since separation becomes larger, thus, in general resulting in a greater accuracy associated with the earlier counts, it is probably worthwhile assigning weighting factors to the various calculated values of CB before the average value is computed. The percent Pr144 contamination in the separated Ce is simply: 100 x (C (C) extrap) 2. Utilize the data obtained for the decay of the Pr'44 activity. If no Cel44 was present, a straight-line semi-log plot should be obtained. The presence of any Ce144 will be evidenced through its daughter activity after the separated Pr activity has decayed away. Determine if any Ce was present in the Pr fraction. See Figure 2. 3. If the two samples are counted on the same shelf, if the samples are counted in an identical manner, and if none of the activity was lost in either sample preparation, then the value of the separated Pr'44 activity, extrapolated back to zero time, plus the calculated value of 0 0 CB should equal (CB)extrap. 6 29

CB + CA if = counting efficiencies 100 Sample 1 Ce144 = 100 10 (Pr144)o = 5 Figure 1 5 c C 1 0 ime (min) 100 Sampl e 2 Pr144 = 95 10 Ce'41 = 1 Figure 2 C Cel44 counted as Pr'44 activity \ WY 0.1 0 50 100 10(m time (min)

6.9 Experiment No. 5 Isotope Dilution Principle: This technique utilizes the very simple principle of proprotional dilution. That is, in mixing two isotopes - (both in the same chemical state) - the relative amount of each in the final homogeneous mixture is proportional to the amount added. This principle is most useful since the relative amounts in the final mixture can be determined for a nonquantitatively separated sample. This method, therefore, is applicable to quantitative determination problems in which quantitative separations are impossibly difficult, very slow, etc. The points needing special attention are: 1) the radioactive isotope used must be in The Same Chemical State as in the unknown, 2) the specific activity (counts per gm) of the known isotope mixture must be known, 3) the separations of the homogeneous mixture need not be quantitative and 4) the specific activity of a portion of the homogeneous mixture (see above, point 2) must be measured accurately. Example: As an example of a quantitative problem to which isotope dilution is applicable consider the following case. A sample is known to contain a minute amount of sulfate and it is desired to know the specific quantity. Precipitation as the Barium or Strontium compound has proven unsatisfactory. Determine the amount of sulfate present by isotope dilution. Method: Mix a given amount of labelled H2S04 to the solution of the unknown mixture. The amount used must be adjusted so that the dilution affect is observable. It is suggested in Overman and Clark page 421, that duplicate samples be prepared and analyzed in the following manner: 1) use two 40 ml centrifuge tubes. 2) accurately pipet into each a "sample of the unknown in the range of 2 to 5 ml". 3) also, add to each tube an accurate aliquot of the labelled sulfuric acid of the size range 2 to 5 ml. (the sulfuric acid is assumed to be 0.110 M and have a counting rate of such a value that the rate is reduced one half on dilution with a reasonable aliquot as suggested above. 6o31

-2 -4) add 10 ml of "distilled" water and heat samples in water bath. 5) add slowly with stirring a slight excess of BaC12 (complete precipitation is not essential) 6) digest ppts for a few minutes then filter through similar weighed filter mounts and dry.. 7) measure the specific activity of the ppts - adjusting counting rate for sample size (self absorption is a major problem with weak beta emitters like S35 and C14) See experiment 6-3, page 240 Overman and Clark for methods for determining self absorption corrections. 8) Calculate the gms of sulfate in unknown.

6.10 DETERMINATION OF PERCENT HALOGEN BY ACTIVATION ANALYSIS Activation analysis, a relatively new analytical technique, provides, in many instances, a very sensitive method for determining, qualitatively and quantitatively, the nature of unknown materials. The procedure, briefly, is to irradiate the unknown material, normally by bombardment with neutrons. Usually,one or more of the various elements contained in the sample will become radioactive. The types of radioactive elements present may then be determined by ascertaining either the half-lifes of the decaying elements, the energies of the decay particles emitted by the elements, or a combination of the two. Using proper calibration procedures, the amount of radioactivity, and thus the amount of the element, may be determined. Unknown samples have been prepared containing approximately 104 gm of an ammonium halide. To measure such a small quantity of a salt, the following procedure was used: 0.100 gm of a salt was weighed out and mixed thoroughly with 100.0 gm of sugar; a 0.100 gm sample of this mixture contains 10 4 gm of ammonium halide. Referring to the table below, it is seen that carbon, nitrogen, and oxygen have very small neutron-absorption crosssections compared with the halogens. (The cross-section for '4N(n,p)14C is 1.76 barns; however, the half-life of 14C is so large, and the 14C B- energies so low that negligible 14C counts are detectable.) As a result, only halogen activity is observed. Before irradiating a sample, it is necessary to calculate the duration of the irradiation desired. To perform the calculation, it is necessary to know the nuclear properties of the isotopes (Table I) as well as the neutron flux and the counting efficiency of the scaler and associated Geiger tubes. The reactor will be operating at full power of one million watts, thus producing a thermal neutron flux of about 1.5 X 1012 neutrons per cm2 per sec at the position where the sample will be placed in the reactor. Although no accurate calibrations have been made, and the values vary from isotope to isotope, it can be assumed that between 6.33

-2 -5 and 10 per cent of the radioactive disintegrations will be recorded as counts by the scaler. In general, using the counting setups available, and considering the statistics of radioactivity, it is desirable to have the level of activity such that between 5,000 and 10,000 counts per minute are recorded. In any case, because of coincidence counting corrections, it is undesirable to have the activity level greater than about 40o00 c/m. Thus, considering that some decay will occur during the counting period, it is probably worthwhile to irradiate the sample so that about '25,000C/m are observed when counting is started. Using a 5% counting efficiency, this would require about 500,000 disintegrations per second. Having decided upon the counting rate and having stipulated the reactor neutron flux, only two variables remain. These are the size of the sample (i.e. the number of halogen atoms to be bombarded) and the duration of the irradiation. Since we have designated the sample size as 10-4 gm ammonium halide, the duration of the irradiation is the remaining variable. For convenience this will generally be limited to 1 second to 10 minutes. The average weight of an anion in a 10T4 gm ammonium halide sample will be approximately 8 X 10-5 gm. The isotopic abundance can be assumed to be 58%, the average of 37C1, 7 9Br, and 1271. Similarly, the average cross-section is 5.3 x 1024 cm2, the average half-life, 27.0 minutes, and the average atomic weight 81 amu. Calculation of Irradiation Duration a = 5.3 X-10-24 cm2 = 1.5 X 1012 n/cm2 - sec T =- 27 min 2 d = 5 X 105 N =(4.7X 10 5)(6.023 X 1023)/81 =3.5X< 1017 atoms halogen 6.34

As a first approximation, we may utilize the expression valid for an irradiation of duration less than 1% of T1. 2 dNt naNM(0n2) where n = t/T1 and t is the irradiation time. Thus: 5 X 105 = n(5 X 10-24)>.5X 1017)(1.5 X 1012)(60 sec/min)(0.693) n = 4.6 X 10-3 t = 0.12:5 minm- 7.5 sec Since n<(001, the calculation is valid. Note: for n>0.01 dN* T?t 1 dtN - = N~(1 - -it) = aNn( 1 - 2) PROCEDURE I. Irradiate a sample of sugar-diluted ammonium halide for about 8 seconds in the reactor. Bring the sample, in lead carrying pig, to the laboratory hood and cut open the sample. Pour the solid contents in (approximately) 50 ml of H20; stir and dissolve. Fill a counting jacket and count the solution for a period sufficiently long so that it will be possible to ascertain the half-life and thus determine whether the activity is 38C1, 80Br, or 128i. This then serves to identify qualitatatively the unknown halogen. II. Another sample of the unknown as well as a sample of diluted ammonium halide of known per- cent and type of halogen are to be irradiated in the same capsule. Simultaneous irradiation of known and unknown results in the same t and. for both samples.* As a result: (dN*/dt) unk = N unk (dN*/dt) known N known *Note: depending on the relative positions of the two samples in the capsule, the neutron flux, A, may differ by as much as 10 to 20o. 6.35

Thus, it is necessary to determine only the relative counting rates, corrected to the same: moment of time since the irradiation. Certain additional aspects must also be considered. Since the counting rates rather than the disintegration rates are being determined, it is necessary to count both the known and the unknown, on the same scaler using the same solution jacket, the same volume of H20 solvent (50.0 ml), and the same Geiger tube. The per cent halogen in the unknown will be larger than that in the unknowns. Therefore, the known can be counted second. It is necessary to count each solution for more than about 10-15 minutes. In the case of iodine, only one isotope, 1281, is produced. Thus, the half-life is about 25.0 min. For chlorine, 3Cl and 36C1 are both produced. However, >1 count per minute 36C1 activity showed to be observed; thus a 37.5 min half-life would be found. For bromine, however, three isotopes: 80Br (18 min), 8~mBr (4.5 hrs), and 82Br (36 hrs) are formed. The 82Br activity will be negligible. However, the S~mBr activity will result in an overall half-life which is greater than 18 min. It is necessary to correct for the 80mBr activity which will be a function of the time since the end of the irradiation. Therfore, if the halogen is bromine, it is necessary to know, for the mid-point of each count, the time since the end of the irradiation. Given in Table II -s the per cent of the observed activity which is due to 80~mBr as a function of the time since the end of the irradiation. Before correcting (using T1 = 18 min) 2 the known and the unknown bromine activities to a common time, it is necessary to subtract the 80mBr activity. Since the 80mBr is counted only through its 80~Br daughter, no corrections for 80mBr and 80Br relative counting efficiencies are needed. Coincidence correction mata will be available. 6.36

6.11 Table I. Nuclear Properties of Selected Isotopes* Natural Percent Thermal Nuclear Half Life Major Decay Isotope Natural Neutron Reaction of Product Particles; Abundance Cross-sec- Energy in tion Cm2 X Mev 1024 H' 99.985 0.330 n,y Stable H2 0.015 5.7x10-4 n,y 12.26 yrs 0.0180pC'2 98.89 0.0033 n,y Stable... C'3 1.11 0.0007 n,y 5600 yrs 0.1565 -N14 99.635 0.10 n,y Stable 1.76 n,p 5600 yrs 0.156pN'5 0.365 2.4X10-5 n,y 7.36 sec l0o-, 6.1y 016 99.759 ________ 07 0.0137 018 0.204 2.110-4 n,y 29.4 sec 4.5-, 1.4y Na23 100. 0.53 n,y 15.0 hrs 1.45-, 2.7y P31 100. 0.21 n,y 14.3 days 1.7P-,No y S32 95.018 0.00.... S33 0.750 0.0023 n,p 24.4 days 0.255 -S34 4.215 0.26 n,y 87.1 days 0.175 -S36 0.017 0.14 n,y 5.04 min 1.6-, 3.1y C135 75.53 42 n,y 13.2 x 105 yrs 0.71 -0.2 nn,p 87.1 days 0.175 -C137 24.47 0.56 n,y 37.5 min 4.85-, 2.1y Br79 50.54 2.9 ny 4.5 hrs.. 8.5 n,y 18 min 2.05-, 0,6y Br81 49.46 3.1 n,y 35.9 hrs 0o44p-,,87 1127 100. 6.7 n,y 24.98 min 2.l1-, 0.45y * Data from Trilinear Chart of Nuclides by William H. Sullivan, 1957+ revisions; U. S. Government Printing Office, Washington, 25, D. C. 6.37

6.12 Observed Activity vs. Time Table II. Percent of the observed activity that is due to Br80m as a function of time (min.) since the end of the irradiation. Data are valid for irradiation durations of zero to about 20-30 seconds, time time 5.00 3.6 24,00 7.0 6.00 3.7 25.00 7.3 7,00 3.9 26.00 7.5 8.00 4.0 27.00 7.7 9,00 4.2 28.00 8.0 10,00 4.4 29.00 8.2 11.00 4.5 30.00 8.5 12*00 4.7 31.00 8.7 13.00 4.8 32.00 9.0 14.o00 5.0 33.00 9.3 15.00 5.2 34.00 9.5 16.00 5.4 35.00 9.8 17*00 5.6 36,00oo 10. 18,00 5.8 37.00 10,4 19.00 6.o 38.00 10,6 20.00 6*2 39.00 11.0 21.00 6.4 40.00 11.3 22.00 6.6 41.00 11.6 23.00 6,8 42.00 I119 6.38

U.13 RADIOACTIVE DECAY D. E. Hull - Activities of Radioactive Substances in Series Disintegration, J. Phys. Chem. 45, 1305 (1941) t/T -At t/T -At t/T - t t/ - t e e e e o 1.000 0.52 o.69'74 1.54 o.7439 3.80o U.Uyi( 0.01 0.9931 0.54 0.6878 1.56 0.3391 3.85 0.0693 0.02 0.9862 0.56 0.6783 1.58 0.3345 3.90 0.0670 0.03 0.9794 0.58 0.6690 1.60 0.3299 3.95 0.0647 0.04 0.9726 0.60 0.6597 1.62 0.3253 4.00 0.0625 0.05 0.9659 0.62 0.6507 1.64 0.3209 4.10 0.0583 0.06 0.9592 0.64 0.6417 1.66 0.3164 4.20 0.0544 0.07 0.9526 0.66 0.6329 1.68 0.3121 4.30 0.0508 0.03 0.9461 0.68 0.6242 1.70 0.3078 4.40 0.0474 0.09 0.9395 0.70 0.6156 1.75 0.2973 4.50 0.0442 0.10 0.9330 0.72 0.6071 1.80 0.2872 4.60 0.0412 0.11 0.9266 0.74 0.5987 1.-5 0.2774 4.70 0.0385 0.12 0.9202 0.76 0.5905 1.90 0.2679 4.80 0.0359 0.13 0.9136 0.78 0.5824 1.95 0.2588 4.90 0.0335 o0.14 0. 9075 0.80 o.744 2.00 0.2500 5.00oo 0.0312 0.15 0.9013 0.82 0.5664 2.05 0.2415 5.10 0.0292 0.16 0.8950 0.84 0.5586 2.10 0.2333 5.20 0.0272 0.17 0.8888 0.86 0.5509 2.15 0.2253 5.30 0.0254 0.18 0.8827 0.88 0.5434 2.20 0.2176 5.40 0.0237 0.19 0.8766 0.90 0.5359 2.25 0.2102 5.50 0.0221 0.20 0.8705 0.92 0.5285 2 30 0.2031 5.60 0.0206 0.21 0.8645 0.94 0.52:L2 2.35 0.1961 5.70 0.0192 0.22 0.8586 0.96 O.5141 2 40 0.1895 5.80 0.0179 0.23 0.8526 0.98 0.50o7o0 2,45 0.1830 5.90 0.0167 0.24 0.8467 1.00 0.5000 2.50 0.1768 6.00 0.0156 0.25 o.8409 1.02 0.4931 2.55 0.1708 6.20 0.0136 0.26 0.8351 1.04 0.4863 2.60 0.1649 6.40 0.0118 0.27 0.8293 1.06 0.4796 2.65 0.1593 6.60 0.0103 0.28 0.8236 1.08 0.4730 2.'70 0.1539 6.80 0.0090 0.29 0.8179 1.10 0.4665 2.75 0.1487 7.00 0.0078 0.30 0.8122 1.12 0.4601 2.80 0.1436 7.20 o.oo68 0.31 0.8066 1.14 0.4538 2.85 0.1387 7.40 0.0059 0.32 0.8011 1.16 0.4475 2.90 0.1340 7.60 0.0052 0.33 0.7955 1.18 0.4413 2.95 0.1294 7.80 0.0045 0.34 0.7900 1.20 0.4353 3.00 0.1250 8.00 0.0039 0.35 0. 7846 1 22 0.293 3.05 0. 1207o 8.20 0.0034 0.36 0.7792 1.24 0.4234 3.10 o.1166 8.40 0.0030 0.37 0.77378 1.26 0.4175 3.15 0.1127 8.60 0.0026 0.738 o0.764 1.28 0.4118 3.20 0. 1088 8.80 0.0022 0.39 0.76731 1.30 0.4061 3.25 0.1051 9.00 0.0020 0.40 0.7579 1.32 0.4-005 3.30 0.1015 9.20 0.0017 0.41 0.7526 1.34 0.3950 3.35 0.0981 9.40 0.0015 0.42 0.74-74 1.36 0.3896 3.40 0.0948 9.60 0.0013 0.43 0.7423 1.78 0.3842 3.45 0o.o0915 9.80 0 oo0011 0.44 0o.737 1.4-0 0.3789 3.50 0.0884 10.00 0.0010 0.45 0. 7320 1.42 0.3737 3.55 0.08 54 10 50 0.0007 0.46 0.7270 1.44 0.3685 3.60 0.0825 11.00 0.0005 0.47 0.7220 1.46 0.3635 3.65 0.0797 11.50 0.0004 o.48 0.7170 1 48. 0.3585 3.70 0.0770 12.00 0.0002 0.49 0.7120 1.50 0.3536 3.75 0.0743 13.00 0.0001 0.50 0.7071 1.5 02 0. 3187 t= time dmi:rng w.hichn. radioactive decay has been in progress. T= half life of the ele!Ti:)t. e-ht= fraction of the number of the atoms of the element which w,:ePei present at time zero time which remain at time (t). N= Noe-7t. 6.19

6.14 Fractional Midpoint of Count vs. Duration of Count Defining t as the duration of the count, T as the half-life of the isotope, and n as the ratio: t/T, x X t = moment during the count at which the instantaneous counting-rate equals the overall observed counting-rate. X - l-y log10 L(gl - i 0010 o: J r t 1O _X o.t o o oN o o o oh o 6-c X Figure 406.40 o 0 -H H co 0 0 H -H -p d -4Fig6 ~rl ~ ~ ~ ~ ~ ~ ~ + x Figur 4.4

6.15 Selected Charts ci Activity vs. Time 4oo0 3~00~I ' — CA + C0 100 __ $0 30, Separated 0 1 2 3 Hours Above: CA (200 days) + CB (20 min) equilibrium. C 100, C = 0. At equilibrium, C CA. Below: CA (0.8 hours) - Ca (8 hours) equilibrium. Co 500 C. At equilibrium, C = CA. A 30B B A 1300 CA ' ~ C 1 I 40 4 12 Hours 'Hoursx

Llo00 200.00 80 - 2 3 4 83 4o-.rr C — Hour s Above: A (20 mint.) - B (10 min.) equilibrium. CA w 100, Co = 10. At equilibrium, 0B = 2.00 CA. Below: A (20 min.) -> B (10mm.) equilibrium. 40 C 500 + ~ l cd N 20 A B Above -. -eparated 0 1 2 0 Hours 6.42.,p~

800 4oo H 200 0c22 Above: Br8om (4.5 hr) Br80 (18 min.) equilibrium. Below: A (20 min.) - Br (10 min.) equilibrium. CA = 100, C = 0. At equilibrium, CB = 2.00 CA 60 60C "A _________ 200 II ZCA I ~ 4. -— 6 -_' —_______ ihours Cr=-At C6.07. 2 B B+CA* 101 6.43

Chapter 7;Experiments in Radioisotope Technology By Ro Borcherts, Jo Sickles, and L.oE Brownell Background The.operation of nuclear reactors.such.as those used in the production of plutonium, has produced large quantities of radioactive fission-product waste materials. These materials cannot be used in the atomic-bomb program nor can they be disposed of, as is customary with ordinary industrial waste, that is, they cannot be discharged into streams or rivers, dumped on the ground, or released into the air by burning because of the hazard of contaminating air and water suppl es and food materials with radioactive poisonso As a result, the Atomic Energy Commission has been storing large quantities of these fission products in underground tanks designed to retain this material for many years. The problem of disposing of these fission products will become more important as the reactor program increases. If the fission products could be put to use in industry, what is at present an expensive waste material might be turned into an asset with the saving of taxpayer dollarso The Atomic Energy Commission contracted with various universities to investigate some of these uses. In June, 1951, Michigan and later other universities (1) received cobalt-60 sources from the Brookhaven National Laboratory of the type described by Manowitz (9). The Michigan source (see Fig. 7.1) was installed in the Fission Products Laboratory of -Engineering Research Institute, University of Michigan, and a variety of experiments were conducted and reported to the Commission (2-5). Additional experiments were supported by the Michigan Memorial Phoenix Project. The small cobalt-60 source shown in Fig. 7.1 was a useful laboratory tool in the early experiments, but only a small percentage of radiation from the small source of cobalt-60 can be absorbed by the specimen placed in the radiation chamberO The small internal diameter of about 1 1/2 in. has greatly limited the size of the sample which might be irradiated and has made it difficult or impossible to conduct many experiments. Therefore, a decision was made to secure another gamma-ray source of greater flexibility. In the second source thee desi;gn was modified so that a greater percentage of the radiation would be usable (see Fig. 7.2). In the experiments with irradiated food, it is desirable to place commercial-size cans in the irradiation chamber. It was found that preservatiorn 7.0o1

of food by irradiation invrqves many problems other than sterilization as experienced in the cannirg,industry and also that irradiated food must 'pbe protected: frqm oxidation, dehydration, etc. as in the canning operation. Prior to use of the larger source, it wasn:.necessary to limit all food tests to food packaged in glass test tubes or plastic containers because the small source would not accommodate the smallest size of tin cia. In the experiments on food sterilization the gamma flux must have sufficient strength so that the irradiation time will be short enough-to assure non-spoilage of the irradiated sample before sterilization can be accomplished. With these considerations it was decided that a cylindrical radiation source of at least 3-kilocuries would be required. and that this source should be designed to irradiate samples both outside and inside the cylinder. For added versatility it was decided that this source should be in the form of a number of rods which could. be set-into a cylindrical pattern or into a layer pattern, depending on what was desired. A few comments regarding the efficiency of using gamma radiation sources might be mentionedO As the intensity of the gamma-ray source is increased, it can be 'used more efficiently, since a greater percentage of the radiation field can be used. For example, a 1-curie soirce is pracitcally useless in promoting chemical reactions or sterilizing biological materials because the field is of such low intensity; whereas k$locurie sources can be used for these purposes. Figure 7.2 shows a cutaway or phantom view and Figo 7.3 shews a plan of the radiation cave. Essential features are the 4-ft. thick concrete walls necessary to shield both laboratory personnel and the surrounding- area from gamma radiation and the 16-ft. well used for shutting off the source. The 4-ft. barrier wall provides a simple labyrinthine entrance and prevents the source from "seeing" the dooro The barrier wall serves to diminish the radiation flux in the labyrinthine entrance so that a heavily shielded door is not required. The door is provided with a mechanical safety interlock, (see Fig. 7.4) making entrance to the radiation cave when the source is in the raised position, impossible. In addition to the safety pro\tfied. by this interlock, a safety light immediately above the locking bar handle serves t, indicate the position of the 10,000-curie source. This device is activated mechanically by the rods of the source itself. As the source is raised to its uppermost position, it contacts a vertical rod which is raised by the final travel of the source and which through a cable indicates the position of the source. Therefore, these two different mechanisms operate independently. These safetyA:measures are supplemented by rigorous monitoring with instruments upon entry into the cave. 7.02

23" LEAD PLUG TOP OF LEAD CONTAINER F.36- ~ ~ ~ ~ ~ ~ ~.I/4. INLE MlIRROR\ B. "AVEO1000 CURIIESN EXPER~IiiBOTTOM OF LEAD CONTAINER 1 /4" STAINLESS ' STEEL JACKET D RAIN LINE Figure 7.1 Small cobalt-60 source MIRROR> 'CAVE' FOR RADIATION 4'CONCRETE SHIELDING;:Fie72Ctwyepeteveofrdainave Xi~~; Jl lY~'DE;:07:OW -hiI+ordr r; sasoIRCE WHENe a WELL-O'DEE

Figure 7~5 shows the loading of the cobalt rods into the aluminum rack by use of special tongs under 16 feet of water. The steel elevator rods, elevator platform supporting the rack, and the underwater light, are shown in the wello Part of the water circulation system required to maintain clarity and control the pH is shown at the lefto After several years of use the aluminum jackets on the rods corroded resulting in cobalt-60 leakage into the water of the well. This required enclosure of the rods and holder in a stainless steel annular container in 1960. -Safety Considerations A number of the experiments described in this chapter involve the use of the large gamma radiation source, and the handling of radioactive materials with activities up to 1 curie. Because of the potential danger involved in working with radioactive material and large gamma radiation sources, a definite awareness and respect should be used when working with these materials. The following paragraphs attempt to define some of the precautions to be used with radioactive materials and radioactive sources (generally sealed radioactive materials)o (S;ee Chapter 1 and 2 of "Radioisotope Technology" entitled "Safety in Work with Radioisotopes," and "Design and Use of Radiation Laboratories"' respectively. Large Co6 Source (61500 curies as of June 1961) The large source in the Fission Products Laboratory in the form of a cylinder has a dose rate of 110,000 r/hr in the center well as of June 1961. The cave, holding the source, has a 4'-0" concrete (solid) wall surrounding it and a labrynth type entrance (see Fig. 708). From the entrance one can see whether the source is up or down. To enter the cave the source is slowly lowered by means of the hand crank into the well of water. (Note the interlock mechanism to prevent the door being opened unless all the cable is unwound). When entering the cave, always carry a monitoring instrument and check it with the Cs137 source inside the door to see if the instrument is working. As far as knwon, lethal doses of radiation in the low dose range cannot be felt, seen, or heard so that you must rely on your instrumento Another, though visible, check can be made by using the mirrors enabling the person entering the cave to see that the source is not in the up positiono Upon leaving the cave be sure to remove the monitoring instrument and turn it off before raising the source. Small Co60 Source (970 curies) Access to the radiation field of the small Co60 source is made by removing the plug at the top of the source. This act creates a rather intense 7003

radiation field above the hole that is conical in shape and should be treated with caution. When using this source be sure to: (see Fig. 7.1 and 7.6) 1o Wear your film badge on the neck level 2. Wear finger tabs 30 Notify personnel upstairs that you are using the source General Procedures 1. Always wear your film badge 2. When working with or around unsealed radioactive material, frequently check your hands and feet by means of a monitor capable of detecting the radiation 3~ Know the properties of the radioisotope and know what you are doing 7,01,.

r~~~~~~~~~~~~~~~~~~~~~ ' -I1 SEln C~wrRTIONX RO OM Ire lWsim a el l LL ab. "~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ —I 01 z3t~w ~'-o j" +~~ t ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~1 o.~ ~ ~ ~ r I, l-0, FLOOR PLAN jr[~~~~~~~~~~~~~~~~~ L '!,. I f I ~~~~~~~~,_ SECT12N. xl-x NOR~TH ELE~VATIONY Figure 73 Plan and elevation sectional views of radiation cave Figure 7.3 Plan and elevation sectional views of radiation cave~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Figure 7.5 Loading cobalt-60 rods into holder under 16 feet of water used as shielding ga~~~~~~~ t~RLW HIGH LEVEL LELEad MECHANNISM LABYRINTHINE ENTRANCE 0,0 CREOS 4 SR 3D00R TO CAVE P~~~~~~~~T ~ ~ 1 DLOOR ONPLCEOP~~ NABLE LV LILLA ASSEN____ PLAN VIEW OF SAFETY MECHANISM FOR 10.000 CURIE COBALT O SOURCE I. 00CR IN WUNLCKED POSITION I. WINCH IN LOCKED POSITION Figure 7.4 Door interlock to radiation cave

0.06 015 0.20 SECOrvD FLOOR 0 ' | # 3. 77 / /' /0/ s I //zO 0.20 0.6_ 3/2 BOTTOM FLOO R \ / 00cobalt-60 sour ce UNITS: MIbJLLlROCNrGErNS PLR MIN UTE 7-OTAL RADIATION/ - - --— JOFT BErq RAIDIATION El-MINATED DArA of 7 4ULY SiL // cobalt-60 source

I0,000 CURIE POINT SOURCE -CALGULATED, NO ABSORPTION X 10,000 CURIE SHEET SOURCE I| ~~~~~ L" - e / -GACALULATED, NO ABSORPTION oV FERROUS OXIDATION MEASUREO c Q \ / MENTS, 13 MARGH'53 U_ = 0 / O FERROUS OXIDATION MEASURE_ X \,\\ / MENTS, 16 MARCH'53 1c _. -A FERROUS OXIDATION MEASUREDISz T A rnC s/ EMENTS, 9 JUNE '53 <j nc + VIGTOREEN RATE METER Mt z_ ~! t\/ / MEASUREMENTS, MARCH'53 o 1,-O VICTOREEN R METER LO I0 I00 -CALCULATED, NO ABSORPTION - 1,000 CURIE ANNULAR SOURCE -CALGULATED,' NO ABSORPTION x 1,ooo CURIE SHEET SOURCE /\-CALULATED, NO ABSORPTION I c___~ ~ 1,000 CURIE ANNULAR SOURCE -CALCULATED, WITH ABSORPTION — O VIGTOREEN R METER MEASUREh; 105 /'t I '' MENT, MAY '52.~ ~ ' VICTOREEN R METER MEASURE-.- MENT, MAY '53 w + V FERROUS OXIDATION MEASUREa- MENT, MAY '52 o FERROUS OXIDATION MEASUREw9~ -ACLMENT, MA O '5 ASU. — \ - -XVITOREEN RATE METER MEASUREIO/o -O IOO DISTAN \E ABOVE MID-PLANE, Al. Figure 7.8 Dose rate on is of -kc Co60 source (4)

40 - - - - I I _ _ — WHEAT (B LK) ' w 10 FLOUR(BULK). o 9 7 - 8 I I I POTATOES 6 I. HALF VALUE THICKNESS OF BULK FLOUR 5 - 1. 16 " POTATOES.I -. "i "S "I S" 66 WHEAT 4 __ __ I I, I I IWHEAT -0 2 4 6 8 10 12 14 16 18 20 22 THICKNESS - INCHES Fig. 7.9 "Broad-beam" absorption measurements showing "half-value thickness" of three foods (measured in the radiation cave, Fission Products Laboratory, University of Michigan).

OPTICAL DENSITY ON BECKMAN MODEL DU METER 0o '- _ _ _ _ 0 H 0 H. 0 H o' F-J- ~ - c+ 0 PI) o ) 0 o 0 F-J0 0 m (D H 0 o pj CD +_ O_ td Pi i

7.1 Experiment No. 1 Measurement of Radiation Field of Small and Large Cobalt-60 Sources Using Various Instruments Operated by Ionization of Gases Discussion Before commencing work with cohalt-60 sources, a basic understanding of the danger of radiation exposure and the procedures for proper use of monitoring instruments are required. See Chapter 1 and 2 of "Radioisotope Technology" and Chapter 11 of "Radiation Uses in Industry and Science" or equivalent. Large Source: Since the large Co6 source is finite in size it is expected that the dose rate will deviate from a 1/R2 curve close to the source. However, as the distance away from the source is increased the curve should approach a 1/R2 curve. Near walls and other objects the radiation field should deviate slightly from the expected due to scattering from these objects. Small Source: With the plug removed from the access hole to the small source, the radiation field along ~the axis of the hole is expected to decrease as the distance along the axis is increased. Because of the effect of sourceshielding geometry, this decrease will be greater than for a point source. Procedure Large Source: The probe of the Victoreen Rate Meter is set up inside the cave with the meter outside. For each probe position change, the source must be fully raised and lowered. At various distances from the source and on the mid plane of the source the radiation field should be measured. Since the cylindrical geometry of the source is expected to give the most intense and most uniform field in the center, be sure to measure both across the midplane and along the vertical axis for the center well position. To check the several probes against each other, include an overlap region of measurements for each probe. Note: Allow five minutes for the Victoreen to completely warm up. Zero the instrument on the 3R scale every time a probe is changed. Do not exceed the maximum range of the instrument when raising the source. Small Source: As the probe of the Victoreen is lowered into the small source, its position is remotely recorded with a meter stick relative to some position, say the top of the small source. When the radiation field fails to register on the Victoreen the Juno Ionization Chamber should be used. Take readings with the Juno meter across the cone of radiation at several distances above the top of the source. For the highest distances note the dip of the radiation field at the axis. When plotting the data, make iso7.11

dose curves, ioe.e curves of constand radiation fluxo Care should be taken to, minimize the time spent in the field surrounding the small sourceo Wear film badges on the collars of your shirts'and wear a finger tab when holding or adjusting the probe in the radiation fieldo AEC Maximum Permissible Doses --- —300 mr/wk —whmole body 1500 mr/wk —hands and forearms (See Chapter 1 of "Radioisotope Technology.") Questions, 1o Why does the radiation field along the center line of the small source fall off faster than a 1/R2 curve? (See Fig. 7.6 and 708) 2. Explain why the radiation field above the small source exhibits a tardioid shape (See Fig. 7.6) 3. ao From the dose rate measurement at the geometrical center of the source, the dimensions of the source, and from the assumption that the source is a cylindrical shell of zero thickness and no self absorption, calculate the strength of both sources in curieso b.o From the data of the 1/R2 region of the large source compute the "point source strength" of the source. c. How do your results compare with the given values of source activity? Large Source 2645 curies - Aug. 1955 Small Source 300 curies - July 1951 (See Figs. 7.7 and 7.8) 4o Explain the principles of operation of the Juno Ionization Chamber and the Victoreen Rate Meter Additional References lo Brownell,._LoEo "Radiation Uses in Industry and Science" (Chap. 11) G.P.O.., June 1961 2. NBS Handbook 59 and Addendum, "Permissible Dose From External Sources of Ionizing Radiation" and "Maximum Permissible Radiation Exposures to Man," September 24, 1954 and April 15, 1958, U.S. Government Printing Office. 30 NBS Handbook 62, "Report of the International Commission on Radiological Units and Measurements," April 10, 1957, U.S. G P O. 4, Lewis, JoG o. et alo, "Analysis of Radiation Fields of Two GammaRadiation Sources," Nucleonics 12, No, 1, 40, 1954 7a12

cu 4-J "3 cli 0 c) "3 0):ji~:~- ~i:~::~:8':j;-:,i:::;:i::::;-::-0 I:::ii.:::;i-::::i:::::ii~l::::i:;i::~i~-;::::::::::-: i::::l:,il::_ii-:`i~i ---:-::::01

702 Experiment Noo 2 Ao'Chemical Dosimetry and Dye (Film) Dosimetry The subject of dosimetry is covered in Chapter 3 of "Radioisotope Technology~" This chapter or its equivalent (4) should be read before conducting this experiment. For a background on radiation chemistry see Chapter 7 in "Radiation Uses in Industry and Science"(1)o Intermediate Field Dosimetry Ferrous Sulphate nDiscuss ion The operation of a gamma irradiation facility or laboratory requires the use of a standard method of measuring radiation. Generally the standard used is the ferrous sulphate dosimeter since it lends itself to wide ranges in dose rates, radiation times, and easily results even though it is a secondary standard itselfo Upon absorbing radiation energy2 the ferrous ions are converted to ferric ions which are easily detected by the rather broad resonance absorption of light having a wavelength of 3040 Angstroms. The mechanism of this conversion is discussed more fully in the article by Weiss (2)o Procedure The intensity of the radiation fields measured in Experiment No. 1 will be checked by ferrous sulphate dosimetryo The dosimetric solution used in this experiment is: (see Refo 3) 1. 2 gm Fe S04 -7 H20, or Fe (NH4)2 (S4O)2 6 H20 2. 003 gm NaCl 3o 110 ml (95-98%) H2S04 (analytical reagent grade) in sufficient distilled water to make 5 liters of solutiono A small quantity of the prepared solution is poured into small vials (about 3 ml per sample) and the vials are then placed in both the center well and the area adjacent to the center well of the large source~ Each vial should be exposed to approximately 5-35 Kilorads. By using the data from Experiment No. 1, the approximate time for exposure can be determined. The optical d.ensities of the irradiated samples are then read on the Beckman spectrophotometer and the relative change in optical density is converted to kilorads by using calibration curve of the spectrophotometer (see Fig. 7.10). For the Beckman use a setting of 304 mu and a slit width of 0.5 mmo 7021

Questions lo Why should the dose to the vials be kept below 40,000 rad? 2. Compare the data from the ferrous sulphate dosimetry to that obtained from the Victoreen in Experiment 1 and attempt to explain any discrepancieso References 1. Brownell, Lo E. "Radiation Uses in Industry and Science," US Government Printing Office, June, 1961 2. Weiss, "Chemical Dosimetry Using Ferrous & Ceric Sulphates," Nucleonics, July, 1952 3~ Weiss, Allen & Schwarz, 1955, Geneva Conference, 14, p. 179 4. "Measuring Large Radiation Doses," Nucleonics, 17, 10, P. 58-78, 1959 B. High Field Dosimetry —Blue Cellophane and the Bragg Gray Effect The small dose limitation of the ferrous sulphate technique has led to the search for an effective dosimeter at high doses. Henley (Ref. 1 and 2) discovered that commercial blue cellophane identified as duPont No. 300 MSC (4) containing a dimethoxy-diphenyl-disazo-bis 8 amino-l-napthol-5, 7-disulphonic acid dye gives satisfactory results for a dosimeter in the megarad region. The dye in the film is decolorized as a result of reduction by radiation in a statistically random manner. A given fraction of the dye is reduced to the lenco form following a first order chemical reaction and the change in color is found by using the Beckman spectrophotometer. In Henley's technique, the strips of cellophane are sandwiched between polyethylene sheets and the resulting secondary electrons from the polyethylene cause the change in color of the blue cellophaneo Having a fixed thickness of polyethylene provides a uniform production of secondary electrons. This production of ionization or ionization effects by secondary electrons is known as the Bragg-Gray EfIect. (See Chapter 2 of "Radiation Uses in Industry and Science.") Since the lightening of the blue cellophane depends on the number of secondary electrons and the number of secondary electrons is dependent upon the thickness of polyethylene (up to a limit), it is possible to observe the effect of thickness on the cellophane by use of different thickness of polyethylene in the dosimetry measurementso An approximate theoretical analysis can be made in the following mannero 7,22

Let Io be the number of gamma photons striking the face of the polyethylene of thickness t per unit area per unit time. Then the number of primary gamma photons at a distance x (O < x < t) in the polyethylene is given by 0 = oe-~lX where [ is the narrow beam attenuation coefficient of polyethylene. (See Chapter 4 of "Radioisotope Technologyt") The number of secondary electrons produced between x and x + dx is Idl = 0oe dx (7.2) which is the number of gamma photons that have interacted between x and x + dx. If, as an approximation, we assume that these secondary electrons are attentuated exponentially with an attenuation coefficient, a, then the number of electrons produced between x and x dx that reach the blue cellophane is dN = IdlIe ( = o(ex)(a(tx)) dx (7.3) where t is the thickness of the polyethylene attached to the blue cellophane. The total number of electrons that reach the blue cellophane is then, N [ Xo (e-x)(et )] dx (7.4) 0 and integrating, I -ft -crt N(t) = A [e -e ] (7 5) The plot of Equation 7.5 will have the following shape: N(t) o topt Figure. 7.11 Sketch Showing General Shape of Curve for Maximum BraggGray Effect as Determined by Blue Cellophane Dosimetry. 7.2.3

The approximate value of the optimum thickness can be found by setting dt topt = 1 lnTye 1 (7e6) From the experimental data the optimum thickness can be obtained and from the literature, I. The resulting transcendental equation (7.6) can be solved to determine the effective attenuation coefficient of the secondary electrons, Procedure Sandwich the strips of blue cellophane between the 1/4" pieces of polytheylene and expose to a radiation field from 1 to 100 megarado Have at least 4 samples in the 1 to 10 megarad region and 4 samples in the 10 to 100 megarad region. Measure the transmission of the blue cellophane at 6550 Angstroms with a slit width of.15 (Note that at this wavelength the tungsten bulb is used as the light source). To obtain optimum sensitivity set the Beckman so that the blue cellophane with largest dose corresponds to 100% transmission. For the samples with a small change use the expanded scale i.oeo, 0ol scale. Plot the data as: %T sample - %T control versus %T control the dose on both linear and semilog paper. Bragg Gray Effect For the Bragg Gray data sandwich the blue cellophane between varying thicknesses of polyethylene and expose for the same dose (approximately 10 megarads). Since the changes in color for the different thicknesses are expected to be slight, use the expanded scale of the Beckman. Questions 1. For low doses Henley gives the equation Dose (megarads) =.68 (T-To)+-C as representative of his data. What equation do you get for low doses? for high doses? 2. From the apparent relation %T = 100e-r r is the optical density given and XT is the percent transmission given by the Beckman spectrophotometer) show that the number %Ts5_ Tc is independent of the initial setting. %Tc 72*.4

Note that difference in optical density of two samples is independent of its initial or zero reading. 3o From the Bragg Gray data compute C --- —the effective attenuation coefficient for the secondary electrons. References 1. Henley, Eo J-, "G a Ray Dosimetry with Cellophane Dye Systems," Nucleonics 12, 9, 62, 1954 2. Henley E. J,, and Richman, D., "Cellophane-Dye Dosimeter for 105 to 107 Roentgen Range," Analytical Chemistry, 28, 1580, 1956 3o Evans, Ro D.o9 "The Atomic Nucleus," McGraw-Hill, New York, 1955 Low Field Dosimetry-Chloral Hydrate (0-1000 rad) Certain chlorinated organics release HC1 when subjected to gamma irradiationo Acid indicators in such a solution, or the testing of the conductivity of irradiated solutions can give a measure of the radiation absorbedo Chloral. hydrate, CC13CH(OH)2 is an organic compound which is available commercially in crystalline form~ It has a molecular weight of 165.42 and a solubility in water of 4,70 gms/100 mi. When irradiated while in a freshly prepared aqueous solution, the pH of the solution decreases in a linear manner with dosage over the range of 0 to 1000 rad. Procedure IT this experliment the pH of the solution will be measured to determine its acidityo Prior to irradiation, prepare several chlorate solutions of different molarity (0o1M to 5M) of at least l00ml volume. Then irradiate small vials of the solutions in ~kown teiLds lor clefinite times. After checking the pH meter against the standard buffer solutions, measure the pH of each of the irradiated vials as well as a control. Questions 1o What is pH factor? 20 Is there any change in the slope of the curve as a function of molarity? How do you explain this result? 7~25

3. From the data closest to the LM solution compute the average energy absorbed to liberate one hydrogen ion. References 1. Hilsenrod, J. Chem. Phys. 24, 917, 1956 2. Woods, R. J. and Spinks J. W. T.,, "The Action of Co60 Gamma Rays and of Fenton's Reagent on Aqueous Bromal Hydrate Solution," Can. J. of Chem. 1475-86, 1957 7~26

7o3 Experiment Noo 3 Experimental Determination of "Broad Beam" Attenuation of Gamma Radiation Discussion The practical calculation of gamma ray attenuation in shielding materials is simplified by the use of a "build-up" factor. The significance of "build-up" factors and the various types of build-up factors used are discussed in Chapter 4 "Shielding Calculations for Gamma Radiation" of "Radioisotope Technology" (1)o There are at least twelve processes by which gamma radiation interacts with matter, but only three, the photoelectric effect, the Compton effect, and pair production are of primary importance (2)~ Of the three processes only the photoelectric process can be considered to represent complete absorption of gamma photons at or near the point of interaction. In this process all the energy of the gamma photon interacting is used to eject an electron from the electron cloud surrounding the absorber atom involved in the interactiono Pair production occurs only with high energy photons such as those produced in the core of a nuclear reactor. The interaction of these photons with absorber atoms results in the production of a pair, an electron, and a positron near the nucleus. of the absorber atom involved in the interaction. The positron has a short life and is annhilated by interaction with an electron to produce two gamma photons at 180 degrees each having an energy of 0o51 Mevo Since the energy of these secondary photons is considerably less than the energy of the initial photon, they will be absorbed in a shorter distance thereby approximating true absorptiono The major problem in the calculation of attenuation of gamma radiation occurs with the Compton interaction. In this case the primary gamma photon interacts with an electron in the electron cloud around the absorber atom, ejecting the electron, but all the energy cannot be transfered to the electron. The surplus energy is expended in the production of a secondary photon of lesser energy than the primary photon and scattered at an angle from 0 to 180 degrees from the path of the initial photono In; some interactions in which only a small amount of energy is expended in ejecting the electron, the secondary photon may:L`be~- almost as energetic as the primary incident photon resulting in subsequent interactions similar to the first. Thus, at the point of interaction the Compton effect results in a production of secondary photons without a decrease in the photon flux, but only a decrease in the average photon energy. The secondary electrons also produce x-ray photons by interaction with the absorbero Thus, there can be a "build-up' in the number of photons and in the number of interactions per unit distance. It is the purpose of this experiment to determine experimentally the "buildup" factor for some common materials. 7o31

The number of primary gamma photons which emerge from an absorber is given by the expotential equation I = (0 e_,lt where 4 equals the initial gamma photon flux, J equals the "narrow beam" attenuation coefficient for the absorber and t is the absorber thickness. In the case of Compton interaction, the above equation may be modified to include the build-up of secondary radiation as follows f (N) = Bf f '(N0) where f (No) is the number of uncollided photons~ Bf is the build-up factor that gives the correct answer f (N) or for most cases 4 = Bf 4o e-it It is important to note that not only does the build-up factor (Bf) change for each f (N0), but it is also dependent on the geometry and atomic number of the absorbero IT the radiation present at a point in an absorber is to be considered, the build-up of radiation due to Compton electrons, x-rays, electrons and annihilation quanta from pair production in addition to scattered photons are involved in Bfo The build-up factor can also be incorporated in the exponential kernel of the equation giving 4 = o e-(bf ))t or 4 40 east where ia is now the broad beam absorption coefficient and is not a constant unless Bf varies exponentially. Procedure The Victoreen ratemeter is placed in a fixed position in the radiation fieldo The field is determined with no intervening test medium between the source and the thimble chambero Subsequent measurements of radiation level are made with successively increasing thickness of the absorbing medium interposed between the source and the chambero The build-up factors and the broad beam absorption coefficients should be shown as functions of absorber thickness. To obtain data for calculating the effective density of the concrete shielding walls of the FPL cave, measurements of the radiation level at each of the outside faces of the cave are taken with the geiger tube instru7032

ment. All measurements are made with the Cobalt-60 radiation source in the up positiono The source is then lowered into the well and the same measurements repeated to determine background. Using the strength of the source computed in the first experiment, the known shielding configuration, and the build-up factors in concrete, apply the data to calculate the density of the concreteo Note any simplifying assumptions made. Figure 7.13 shows some half-value thickness for broad beam absorption in three food substances. List in the results and compare your half-value thicknesses for each of the absorbers tested with those in Figure 7.13. Questions lo With increasing thickness of the denser materials such as lead and concrete you should observe a deviation in the slope of the curve with the larger thicknesses. What is the reason for this change? 2. With the less dense materials a plot of log of percent transmission versus thickness may show a slight increase before leveling off to essentially a straight line. Explain the reason for this phenomenon. References lo Brownell, L. Eo "Manual for AEC-NSF Institute on Radioisotope Technology," June 1961. 2. Goldstein, H. "The Attenuation of Gamma Rays and Neutrons in Reactor Shields," USAEC, U.S. Government Printing Office 2, May 1, 1957. 3. Price, Horton, & Spinney, "Radiation Shielding." 4. Rockwell, T. "Reactor Shielding Design Manual." 7~33

P-H P- H vs Dose 0.2 m/t Solution 50 R/Min Dose Rate 0 5 10 DOSE ( I00 RAD) Figure 7.12. Experimental Observations of the Change of Ph of Freshly Prepared Chloral Hydrate Solution vs. Dosoge of Gamma Radiation. (Unpublished work, J. F. Rice, University of Michigan. )

40: 1 ~ - ~ I~ t <iTHEA T(BULK) o w 10 o I _ FLOUR (BULK). 09I 9I r_ 7t ~ ~ ~ J.!POTATOES (BULK), 6 | I -HALF VALUE THICKNESS OF BULK FLOUR S' r I J yr | " "I I " " POTATOES I -.,..s,... WHEAT 4 - - I I I I I I I I 04 2 4 6 8 10 12 14 16 18 20 22 THICKNESS - INCHES Fig. 7.13 Absorption measurements performed in the radiation "cave" at the Fission Products Laboratory.

7.4 Experiment No. 4 Radiography with a Source-Target Mixture Discussion There has been a real need for a rugged, portable X-ray facility for field radiography when it is not possible to transport the patient to a hospital where conventional X-ray equipment is available. Gamma radiation or the "bremsstrahlung" from various beta-particle sources offer a possible solution to this problem in the form of a portable radioisotopic sources. Studies of a promethium-147 tungstate source-target mixture for use as an X-ray source were reported by E. W. Coleman,* L. E. Brownell,** and C. J. Fox*** at the 1958 Geneva Conference on Peaceful Uses of Atomic Energy (1). The successful application of radioisotopes in medical radiography requires an abundant supply of source material meeting the exacting requirements of the art. An isotope suitable for medical radiography must have a high specific activity so that sources of small diameter and high radiation intensity can be made. 'The radioisotope should emit radiation having essentially the same characteristics as the radiation from an X-ray tube used for diagnosis. To obtain a differential absorption between various body tissues, radiation is required with energies in the range of from 30 to 100 kev and preferably from 30 to 80 kev. The half-life of the radioisotope should be long enough to permit use of the source for a reasonable length of time before replacement is required. Partable X-ray units utilizing radioactive sources of radiation have been reported by the Army Medical Research Laboratory. 2-6 While such units have certain advantages over conventional field radiographic equipment, the quality of radiographs produced is not considered acceptable. The presence of even a small amount of higher-energy radiation results in a loss of radiographic contrast between bone, muscle, and fat. Thulium-170 was found to be a usable radiographic source, but because of high-energy bremsstrahlung originating within the source, the radiographic quality was poor. Radiographic sources using pure beta-emitting radioisotopes and external target foils have been proposed. The disadvantage is the limitation of source thickness to the depth of penetration of toe beta particles in the source material, since all greater thicknesses will result in self-absorption *Picker Research Center, Picker X-Ray Corporation, Cleveland Ohio (formerly with Fission Products Laboratory, The University of Mich., Ann Arbor, Mich. **Professor of Nuclear and Chemical Engineering and Supervisor, Fission Products Laboratory, The University of Michigan, Ann Arbor, Michigan. ***Research Assistant, Fission Products Laboratory, The University of Mich., Ann Arbor, Michigan 7.4.1

of beta particles in the source rather than in the target material. This is a severe restriction with sources of long half-life such as strontiun-90, because the maximum specific activity of such sources is limited by their slow rate of decay. This means that if a small "point" source is used, the exposure times will be very long, and if a source of larger area is used, the X-rays will not be well collimated and a sharp radiograph will not be possible. Another disadvantage is that the electromagnetic radiation produced by self-absorption of beta particles in the source and from backscatter in the shield may not have the energy spectrum necessary for a good radiographo Coleman and Brownell 7,8 proposed to avoid these difficulties by use of source-target mixtures. If the target is intimately mixed with the beta-particle source, the source thickness may be increased significantly because the electromagnetic radiation is more penetrating than the beta particles. Furthermore, if the atoms of the beta-particle source are chemically combined with the target atoms, each source atom will be surrounded by target atoms. This increases the probability of interaction between the beta particles and electrons surrounding the' target atoms as compared to electrons surrounding other source atoms. In view of logistic problems expected in storing, transporting, and replacing sources of short half-life, such as thulium-170 (127 days), only those isotopes with a half-life of one year of longer should be considered satisfactory. A "use-life" of three to four half-lives was considered reasonable, based on operating experience with thulium-170. Because of the low specific activity of strontium-90 and americium-241, these radioisotopes were not considered satisfactory, and the authors limited their sutdies to promethium147 and thallium-204. Since the promethium-147 sources were found to be superior to the thallium-204 sources, only the promethium sources Vill be described. A 1-curie promethium tungstate source was ordered from the Oak Ridge National Laboratory. The preparation of the tungstate and the encapsulation were performed by Mr. R. So Pressley of the ORNL staff. This source was encapsulated in a modified Standard Oak Ridge capsule. The modification consisted of pressfitting a 4-mm-ID aluminum sleeve into the source cavity and machining the stainless-steel window to 0.005 in. Aluminum was chosen for the sleeve to minimize bremsstrahlung and characteristic X-ray production in the vertical walls of the source cavity, so that the effective focal spot size would be 4 mm. The capsule was silver-soldered after filling and placed in a larger lucite capsule with an air-path aperture to the window. The plastic capsule was used only to eliminate the possibility of accidental contact with the thin window during handling. The question of mechanical strength in a 0.005-in. stainless-steel window., and the relatively high absorption factor for low-energy radiation (e-UX = 0.25 at 25 kev) suggests that a thicker beryllium metal window:2.would be stronger and should improve the efficiency by attenuating a smaller part of the low-energy radiation. 7.4.2

Comparison of Spectra From Experimental Sources In Figo 7o14 the spectra of three possible radiographic sources (thulium170, thallium-204 iodide, and promethiumn-147 tungstate) are shown and may be comparedo Although all three sources have peaks in the characteristic X-ray region, the broad, flat, higher-energy bremsstrahlung spectra, of the thulium170 and thallium-204 sources indicate that shielding difficulties and poor radiographic contrast.may be expected from these sourceso The spectrum of the promethi.um sources is a definite improvement over thuliu-170 with regard -to higher-energy componentso The major peak is at a higher energy than the thulium, but is still in the diagnostic region. A low-energy peak could be obtained by use of molybdate instead of tungstate as target materialo A radiograph of a hun rand (Fig 7 15) was made with the promethium tungstate. -The original radiograph was well detailed considering the source diameter of 4 mm and the source subject distance The quality of the radiograph was adequate for fracture or foreign-body localization~ The wrist detail was good, and the fingernails are easily recognized in the original radiograph. Current technology indicates exposures of less than 1 hour are possible:. The major improvement over similar radiographs produced with thulium170 is in radiographic contrast. Thulium radiographs exposed for maximum detail are grey and white, instead of the continuous gradation from black to white required for excellent contrasto The promethium radiograph shows improved contrast,.probably due to the absence of-high-energy bremsstrahlung in the spectrum. Technical Radiographs Using Prometh-ilum-147 Tungstate Source The use of promethium.-147 sources for technical radiography is demonstrated in Figo 7.160 A variety of small objects is shown in varying degrees of detail. The detail in the expandable metal watch band is sufficient to determine the integrity of the small spring in each linko The thin mica standoff insulators and the flaments in the electron tubes are easily visible. The lack of good detail in the stopwatch is perhaps the best example of the improved spectrum of promethiu sources relative to thulium sources. Similar.:.,adiographs with thulium have shown complete detail of the watch, indicating the presence of a higher-energy spectrumo As- a result of these studies the conclusion was stated thbt radiographs made With promethium tungstate source-target mixtures show im.proved contrasj&tx-;compared to thulium-170 radiographs. Limited tests with radiographs of.,'3:tomo ercial objects of low.mass indicate the possibility of use of promethium tungstate sources for industrial radiography of light-weight productso However, the efficiency of production of the radiographic image must be in7o43

creased to shorten the required length of exposure before medical application can be realized. Procedure The electromagnetic radiation spectrum of a source-target mixture of Promethium-147 tungstate is to be determined and a radiograph of several articles using this source-target mixture is to be produced. Promethium147 is a pure beta emitter of 0.223 Mev with a half life of 2 7 years. Since it is a pure beta emitter, an intimate mixture of pm147 with a heavy element should provide an electromagnetic radiation spectrum similar to that of an x-ray tube - i.e. a continuous spectrum due to bremsstrahlung superimposed on x-ray peaks that are characteristic of the elements in the mixture. Such a source target mixture might have practical use as a portable x-ray machineo However, one of the major disadvantages of the source-target mixture idea is the long exposure times required to produce a radiograph. The time required to produce a radiograph could be reduced by making a larger volume of the source-target mixture. But this would result in poor definition of the radiograph and as the size of the source is increased, the increasing effect of self absorption gives somewhat of a "maximum effective volume" of the source. The method used to determine the spectrum of the source-target mixture will use a NaI(Tl) crystal scintillation detector and a.pulse height analyzer. The photo peak of a known monoenergetic gamma ray source will be analyzed to correlate the scale of the pulse height analyzer to Mev units. Then the spectrum of the source target mixture of Promethium tungstate will be observed. A radiograph of several small objects will be made which will be observed a week later for the effect of source height distance on definition and contrast. Questions 1. Why is the pulse height analyzer and associated components unable to resolve the characteristic x-ray peaks of the elements in the source-target mixture? What energy should these x-ray peaks be? 2. From the source-target spectrum obtained with the pulse height analyzer it is inferred that it is the actual spectrum of electromagnetic radiation from the source-target mixture. This is not true, for the spectrum is a measure of the energy deposited in the crystalo Why can we assume the spectrum of the source is the same or somewhat near the same as the spectrum we obtained? 7044

3. From the source date (1 curie Pm147 - April, 1957, 2.6 year half life) compute the current required of an x-ray machine to produce a "radiograph" in 1 minute compared to an 8 hour exposure to the Pm147 source. Assume the spectrums of the x-ray machine and the source target mixtures to be the same. 40 An important parameter in the use of a radioisotope is its maximum specific activity (curies/gm or curies/cc). The source used in this experiment is cylindrical in shape having a diameter of 4 mm and a height of 3.5 mm. If the source-target material is pure Pm2 (W04)3 and the promethium is pure Pml47, what is the maximum activity that a source this size could have? (Assume the density of the source target mixture to be the density of tungsten). 7~45

10 TM-170 I I PM 147(W04) 8 _____ TL204I Ui~~~~~~~~i\ 7 ~ ~ ~ r _ I n~~~~~ r~ I ~ I I I _ _ _ _ _ _ _ _ _ _ _ I~~~~~~~ I- ~ ~ ~ ~ ~ ~ ~ ~ EEG I 9' z~ig. 7~14 Comparison of the spectra of three radiographic source00 115 _ Lu~~~~~~~~~~~~~~~ 2 ~ ~ - _ __ o 20 40 60 80 100 120 140 160 180 200 220 240 ENERGY - K EV Fig. 7.14 Comparison of, the spectra of three radiographic sources.

Fig. 7.15 Radiograph of human hand, 79-hr exposure at 20 in. from a premethium tungstate source. Fig. 7.16 Technical radiograph, 44-hr exposure at 10 in. from a promethium tungstate source.

7.5 Experiment No. 5 Area Decont amination Discussion Familiarity with the methods and techniques of low level area decontamination and with the effect of surface texture on the ease of surface decontamination is important to the Health Physicist. (See discussion and illustrations on decontamination in Chapter 2 of "Radioisotope Technology.") Generally the subject of area decontamination can be divided into categories of pre-incident planning, the acute phase, recovery phase, and techniques. In pre-incident planning such items as proper facilities, and laboratory equipment and good experiment design are important as well as the choice of isotopes used. For example the use of 54 day Sr89 is much safer than the use of 28 year Sr90. For a low-level contamination by liquid, absorbant paper, trays, a mop and bucket and shoe covers are essential. For high-level liquid contaminations isolation equipment such as signs and barriers, protective clothing, waste storage containers along with the previously mentioned equipment is needed for the incident. Dusts present additional problems. Respirators, vacuum cleaners, protective clothing, shower facility, isolation equipment and waste containers may be needed to handle the incident;. The acute phase occurs at the time of the incident. The first step may be to evacuate the area to a safe distance, but not to spread possible contamination nor to leave for clean areas until checked for personal conteam.ination. Next, call a Health Physicist for assistance and begin to assemble equipment for decontamination. Rotation of personnel may be necessary if high radiation fields are involved. If radioactive dust is suspected to be present, an air sample should be taken. In the recovery phase take instrumentation into the area that is capable of detecting the contamination. A Geiger tube type of instrument sensitive only to gamma rays is not suitable for detecting alpha or beta particle contamination. In planning and executing decontamination procedures, evaluate. the efficiency of each procedure. In general, decontamination techniques fall into two classes. 1. Isolate the area or object and allows natural decay to reduce levels. 2. Decont aminate 7.5.1

The types -of decontamination can be classified as either 'roug ff to permit limited use or occupation of the area and a "detailed." decontanmination which reduces radioactivity in so far as possible, to a sat;isfactory or low level. The processes of decontamination depend on the type of contamination and the nature of the surface. The easiest surface to decontaminate is a wet contamination on a smooth non-absorbant surface. This can be cleaned by flushing with water and detergents followed by scrubbing and further flushing. Steam under pressure is sometimes helpful. For dust, use vacuum cleaning If further treatment is necessary, brush and vacuum again. On greasy surfaces, remove the greasy material with dry cleaning solvents. If neces~sary, follow this with scouring with water soaps and. detergents. Deeply absorbed contaminations present a problem that may only be disposed. of by removal of the surface while at the same time preventing further penetration of the contaminant. For example, removal of the top few inches of earth, removal of floor covering or paints, abrasion, acids, or special chemical compousnds Firmly held contamination can only be "adecontaminated" by covering the area with a suitable thickness of sealing material, or disposing of the contaminated object or clothing by burial in the earth or at sea in weighted sealed containers. General facts about contamination are that it will not be uniform, smooth surfaces are less susceptible than rough surfaces, cracks or crevices collect it, movement of people and equipment will spread it, and no process will neutralize it. In "rough"' decontamination speed may be an essential consideration and. as a result simplicity is usually required. 'The procedure selected should be rapid, suitable for the material, not require large quantities of special or dangerous chemicals, and make use of available erquipment, services and material. The decontaminating of personnel usually requires successive use of ssoap and water scrubbings, checking for reduction in activity more scri-bbing. Remember to stay out of "clean' areas unless you are "clean'," Figure 7.17 shows the checking of shoe covers for possible contamination before stepping onto a "step-off" pad in the Hanford area, A "poppie" coiunter sensative to alpha particle radiation is being used. Health physics procedures require the employee to check his protecitve clothing before leaving a contaminated area. After checking his shoe covers, he can step onto a step-off pad. Here he removes his outer layer of protective clotheli ing -A gloves, head-cover, coveralls, shoe covers, This procedure p:revents spread. of contamination (by particles) into clean areas. 7.5 2

................ -~;-ii~i~i-i.-.-.iil~-.iiii~iiii Fig- 7.17 Checking shoe covers at the Hanford Plant for contamination before stepping onto step-off pad (Courtesy General Electric Co.) __~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~:~:-::::::::ri:: Iiii'~ —::::. iiii~ i ~i i-~iiii~iii::iiiii _'-'-:: —' * itiiils:':':ii:- ai;i~~ii~ ii:_ - I |~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~B~j:iii~~ii~iiiiiiiiiilii iiiii~~::::: _II| I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~1:::-:.:i:i:: —,:ji~:i~i~-ii: ---:::li:I:::::~~:-I: _:: ii: -:::: X: l:'''I:''':::::::___~'-:_:-:::::: 1::ji i l I: 11-: — a:i: -::::I EA...~~~~~~~~~~~~~~~~~~~~i.-. iii~i' ~iiiiiFig. 7.18 Periscope picture of equipment in decontamination canyon at~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~_:::-::-::::a::~ Hanford~~~~~~~~~~~~~~~~~~~~~~~ Plnt(Cures Gnral letrc o.

kO C) F (D0 ~.......~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.... (DO C) 00 o (D H.::r' 00 0t. 0D. i.. el — I —' 0.. O't: ~ i!i!~!;i;ii:ii;;;ii;;!;!i!;;;ii~iiiiii~iiiii!;iiiiiiiiiiiiiiii F - -......... O iiir~~~~~~~~~~~~~~i:::r-i:::::i~~~~~~~~~~~~~~~~~~~~~~~~~~~i~~~~~i:_:-:i~~~~~~~~~~~~~~~l -,;::i..............-..... "R a~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~....... ~~~~~~~~~~~~~~~~~~~~-/ P~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~..... ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~Z1~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~......... 09~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~............. O~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.... FC........................... (D~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.............. Y~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~.............. Co~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~............. cc,............-jij j:::::li~l:.............::1:n::::i:jj F ji...., ~~~~~~~~~~~~~~~~~~~~~~~~~~~:iii:i::ii:::;:::::!...........: ~~ jj::j?::i::::::::::::::ij:::::::i:::::i:~~~~~~~~~~~~~~~~~~~~~~~~~~~.......................:::::: ~~i ~ ~ ~ji~~~~~l.:::-:-;i:-;i~~~~~~~~i:;i-:1:::-:-:-::-:::r:;:::::;~:.......................... -: j::::::i:::::i C+..................ii~bi -16iii~ 09~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~....................:-::":::i:::.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ W::::::j:jj:li\~i~::~:

Figure 7.18 is a copy of a phqtograph taken through a periscope from the over head crane in a canyon building at Hanford used specifically for decontamination. Expensive stainless steel equipment such as shown in the figure become highly contaminated with fission products during the processing of uranium fuel slugs. This equipment is decontaminated by handling remotely and transfered to canyon areas for disassembly by remote operations. The equipment is then washed by detergents, acids, and water to decrease the decontamination uo a sufficient level uo permit final decontamination and cleaning manually by operators dressed in SWP (Special Work Permit) protective clothing such as used by the worker shown in 7.17. FYig-re 7.19 shows a typical canyon separations plant at Hanford. It is a long, windowless, concrete building, partly underground. The plant separates plutonium from irradiated uranium and recovers the uranium from the mixture of fission products for re-use in The atomic program. Many of the operations inside a canyon are controlled remotely by viewing and handling devices as a means of protecting the workers from radiation. Procedure From the above discussion, plan and conduct a decontamination test for Na-24 from painted and unpainted wood, concrete block, lead brick, linoleum, and or other surfaces. References i. Lane, "Contamination & Decontamination of Laboratory Bench-Top Materials," Nucleonics, August 1953, p. 49. 2. Breslin and Solon, "Fallout Countermeasures for AEC Facilities," NYO 4682-A, December 1955. 3. Curtis, R. L. "Decontamination - A Literature Search," Y-964, May 19, 1953-70 titles are included. 4. Handbook 48, "Control and Removal of Radioactive Contamination in Laboratories," UoS, GPO, December 15, 1951 5. Handbook 69, "Maximum Permissible Body Burdens and Concentrations of Radionuclides in Air and Water for Occupational Exposure,' US GPO, June 5, 1959 7,53

7.6 Experiment No. 6 Treatment of Radioactive Wastes Discussion Familiarity with the methods for decontamination: simple distillation, cation exchange, anion exchange, and combinations of these processes is important. Treating liquid radioactive wastes such as fission products, involves separating or concentrating the radioisotopes as much as possible and then disposing of the concentrate by approved techniques. Figure 7.20 shows a typical reactor area at Hanford where plutonium is produced from natural uranium. Each reactor area contains its own water purification plant to provide cooling water for the reactor and its own steam plant for auxiliary power. The Columbia River is shown in the background. Fuel slugs from these plants are processed in canyon buildings such as shown in Figure 7.19 of Experiment 7.5. After recovery of plutonium and uranium, the fission product wastes are concentrated by evaporation for recovery of nitric acid. The acid solution is then neutralized which results in precipitation of some of the fission products. The slurry is sent to underground storage in large reinforced concrete tanks lined with plain carbon steel. Figure 7.21 shows a picture in the radioactive waste storage area located in the desert at the Hanford works, The four workmen in SWP clothing have emptied one of the storage tanks by pumping to another tank and are about to take photographs for inspection of the tank interior. Infrared film must be used because residual activity from the fission products will fog conventional photograph film. In the ion exchange method of concentration, the radioactive ions are absorbed on the surface of a resin. The resin can be pictured as very large molecules linked and cross linked to form a chain. Protruding from. the chain at regular intervals are groups (generally sulfonic groups for cation regions and amine groups for anion resins) like the spines on a porcupine. When the ions flow over the surface of the resin, the resin group absorbs it and in the case of the cation resin, releases hydrogen. The equilibrium between the hydrogen from cation resin and the cations can be represented by the equation +n + M + n RH n RM + nH where R the resin. As the equation shows the resin can be recharged and 7.6.1

M is the cation with an acid. Recharging with a complexing solution will selectively remove the ionso By this technique the ion exchange method can also be used to separate the fission productso Procedure In the distillation tests, samples of 1 to 5 cc of the distillate should be taken periodically to determine how the decontamination factor varies with the amount of standard solution lefto When setting up the ion exchange apparatus, the resin should be mixed in water and poured into the column. At no time during the filling of the column or filtering the fission product solution through the column, should the meniscus be allowed to drop below the surface of the resin, otherwise air entrainment will occur, lowering the amount of surface of the resin available to ion exchange. Samples should also be taken periodically from the ion exchange column since the first few samples will contain an excessive amount of water. In using the well type scintillation counter, make sure the plastic insert is in the well to avoid a contamination incident. In cleaning glassware use the "hot" sink to wash out the major portion of radioactivity, then use the cold sink to finish the job. Put spent resins and sand in the carton marked "solid waste" and place the liquid heel from the distillation flask in the bottle marked "wastes." Also check the rubber glovese If they are "clean" replace in box, if not, toss them in the "hot" can. In the hot lab everyone is to wear a lab coat and no one smokes, and no one is to leave the lab area without checking their hands and feet for contamination. Questions 1. Is the difference in the counting rates of the standard solution between the beginning and end of the lab period enough to apply a correction to each of the readings? 2. Describe two procedures both of which would give a decontamination factor of 108 Each procedure may involve multiple stepso References 1o Bleuler & Goldsmith, "Experimental Nucleonics," Experiment #12 Rinehart & Co., New York, 1957 7.6.

2. Nachod, Fo C., Ion Exchange, "Theory and Practice," New York, Academic Press Inc., 1949 3. Kunin, Ro, and R. J. Myers,"Ion Exchange Resins," New York, John Wiley & Sons, Inco, 1950 7.63

;::: ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ AA /-'::~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~rp '-V~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~nutilzn

Experiment No. 7 Measurement of Per Cent Moisture by Neutron Slowing Down Discussion A calibration curve for the per cent moisture present in sand is to be obtained by measurement of the thermal flux of neutrons. Neutrons from neutron (a,n) sources and from thermal neutron reactors are born at energies of several Mev lose their energy in collision with other nuclei until they are in thermal equilibrium (0.25 ev) at which time they wander around the medium until they are absorbed or leak out, The source of high energy neutrons in this experiment is from the (a,n) reaction on Beryllium -- the source of alpha particles being Pu. The detector, or the absorber, is a B10O F3 tube (the B10 has a 1/v cross section for thermal neutrons -- see reference 2) and detects via the ionization produced in the (n,a) reaction on A hydrogen nucleus has the same mass a neutron which permits maximum transfer of energy on collision. Therefore, neutrons will require fewer collisions with hydrogen of the water than with silicon or oxygen of the sand (see question 4) to reach the thermal region and be counted with the B10O F5 tube. The count rate from neutrons scattered in a bed of wet sand will be dependent on the number of hydrogen nuclei per unit volume which in turn is proportional to the number of neutrons that reach the thermal energy region - this is the neutron, slowing down density - q. (See Ref. 2 p.57-60.) For a point source of nneutrons in an infinite mediuim 9 obeys the equation where r is the distance from the source to the detector and Tis the Fermi Age E 0D dE T 6 If hydrogen is taken as the only moderator an6-1 the scattering cross section, rVIs, cfronside-,red c-n-onsant, then_:~'

-thus, 3 [in (Et/Eo)] q R N as for which small r gives a cubic dependence on Procedure (J. Sickles) The equipment to be u.sed. iln this expr t ccnsists o e Pu-e source and the following detection equipme0 tube, ir linear amplifier and scaler. A block diagram of the neutronmoisture ae is shown in Fig. 7. 22. BF3 Counter Pre-amp Linear Aplifier Scal Neutron Source Fig. 7.22 1. Set the scaler voltage -t that recommended. f or the particular B F tube 3 you are using, i.e. 1450 volts for the B F5 tube at 20 mm pressure. Also set the internal voltage of -the scaler at 0. 75 volts and. adjust the discriminator setting of the linear am~plif ier to a value -which eliminates amplif ier noise but not neutron pulses. An oscilloscope may be used. for this purpose but in this experiment a discriminator setting, of 10 will likely suffice. 2. Measure the initial moisture weight per c'ent of the sand. and. then detenrmine a calibration curve of moisture weight per- cent vs count rate by taking count rates at intervals of about 3 per cent until saturation is reached.. Use counting times of at least two minutes and establish a a by taking three measurements. at each point. (:See Ref. 2). Caution: Leave the B F tube 'in the probe only when measurements are being made. This will prolong the lif e of the tube. 5. Tes th feto emtyb lcn h proe a seea dIfferent

Het'erences lo Murray, R.L. "Nuclear Reactor Physics, Prentice-Hall, 1957 2. Price, "Nuclear Radiation Detection," McGraw-Hill 1958 (B1OF tubes and assiciated counting system) 3. 3. BNL-325 —Cross Sections for Nuclei ("Barn" Book) Questions 1 How well does the shape of the countin rate curve compare with that predicted by the above theory? Explain any discrepancies 1.bo What is the "mean-free-path" of the neutrons at this vLue of percent moisture? (Consider only the hydrogen) (See Ref 1.) 2. Explain why the counting rate stays relatively constant below this value. 3- Show that the Bl.0F3counter triuly: measures the thermal flux (See Ref. 2.) 4. On the average, how many collisions does it take for the neutron to get from 10 Mev to 4 ev in (a) hydrogen,(b) oxygen, (c) silicon. (See.Ref. 1, p. 36-37.) 5. Sketch the ox'oss section curves for hydrogen,, silicon, and oxygen. 6. Obtain equations 1-4 above.

pendix for Experiment No. 7 Outline of Neutron Fundamentals 1. Possible existence suggested in 1920. 2. Studies in 1932 by Curie-Joliets of secondary radiation produced by alphas on Beryllium. Believed to be new type of electromagnetic radiation. Could not explain ejection of protons from paraffin by new radiation. Protons had a range of 40 cm in air, equivalent to 5 Mev which would require 55 Mev photons. Be9 He4C (Wrong E 4Be9 + 2 + E 6 3 60 + h Where E1 KE of - particle. 9. 01503 + 4. 00388 + 5 Mev - 13. 00751 + hv amu amu (0.00536 amu) 1 amu = 931.2 Mev.. 5 Mev. 00536 amu balance of energy plus mass: AE -nm C hv 0. 01676 amu 15. 6 Mev available. (Not enough f or 55 Mev photon.)

35. Neutron existaace verified by Chadwick in 92 He showed that radiation could not be electromagnetic unless energy-of radiation was considered dependent upon nature of target nucleus. Example Protons from paraffin5.7 Mev., need5 755 Mev. ys Protons from nitrogen-ol. 2 Mev, need 90 Mev s (Photon E must increase with mass of recoil atom), Difficulties disappea if radiation treated as particles of mass and 0 charge Be9 H+ 2e4 6C13* 6C12 +n 1k + 2 -6 -6 + 0 + Q (Correct Equation). 4o. a-n Reactions 5Li + 2 He 5 5 + 4Be9 + 2He4 6C13 61l2 +B2l, 0HeQ4 - 6n N11 2He4 l 8N15*. N14 n 5 7 ~+ 0 +Q 1l9 4.23* 22 1 ~F +2He I-,1Na 1 Na + on Al 27 He4 p~l P50 on 1 13 + 2 -15 -15 +0 Q A lj.ZA+ 2He _> (Z~2)Y(A+~)* Y> z+ AW+3) + 0n Q, uanta,( 2-7.7, 4.6 4=7 + -64, Mev y<s for Be reaction)

5. Chadwick' s Demonstration Apparatus C I D Neutrons W Amplifier Be L Paraffin C ' Evac Chamber D Po a Source Be -.Beryllium Plate, I = Ionization Chamber W Wind-ow 6. Thermal Neutrons Neutrons in temperature equilibrium with their surroundings do not have a fixed speed. but a distribut ion of speed that is accurately characterized by the gas equation of Maxwell n(v) ~ (2" KT) e where no is the number of neutrons per cubic centimeter K (.Boltzmann constant) =.L38 x 10-16 ergs/~ K T is the temperature of the medium OK For room temperature, T = 2935K, the distribution looks as shown on the next page. 7.7,6

The most probable speed is given by the peak of the curve and can be found y setting dn = 0o This speed turns out to be vp = 2KT = 2200 m/sec. dv m or neutrons in a medium characterized by T = 293~K.A more useful quantity desibing the speed distribution is the average speed, V V = fo~ vn (v) dv 00 ---- ~ 2 fo n (v) dv VP for neutrons at T 290, V = 2482 meters/sec )nsider hydrogen atom in heavy water at 253 C Most probable vel, vp = 2200 m/sec oE. = 1/2 Mv2 = 1/2 (1.66 x 10-24 gm/atom) (2.2 x 105 cm/sec)2 - 4.0 (1014) Ergs Ergs) = m (gmso) (C, cm/sec)2 where C = Vel. of light = 2.99790 (10)10 cm/sec (1L-14) ergs 0 (10-) ergsMev = 0.025 ev. Energy of thermal neutron 710o eg Me 0.5 v. -2200 /e - 42KT/m 0,3 0.2 0 _ _ _ 0 1 3

Ep = Energy corresponding to most prob. Vel. = 1/2 m vp2 = KT K = Boltzmann's constant = 1.38 (10-16) erg/ ~ K = (R. ideal gas constant/mole 1 Na, number of particles/mole vp (25~ C) = 2 (1.38) (10-16) /(1l66 x 10-24) = 2~2 (105) cm/sec 7. Neutron Cross Sections Cross section is measure of probability of a process -- --- - Area = a CO0 0 _O~ ~~ ~~Absorption n Incident Neutrons 0 -~^~~~ /, 9,No Collision i ~~ /....Scattering_^ Consider cube 1 cm x 1 cm x 1 cm Based on geometry, probability of collision?Target~ Area =N~ Taget Area =N where N = no. atoms/unit vol. Total Area a / a = target area/atom Geometry concept is approximate with thermal neutrons 7.7.8

Example: Diameter of atomic nucleus-, 10-12cm Area of atomic nucleus.$ 10-24 cm2 Experimental cross section of U235 - 650 (102) cm2 Experimental values are used. Holloway and Baker (Los Alamos) suggests Barn as unit: 1 Barn = 1024 Thermal Neutron Microscopic Cross-Section, a Barns Isotope o H 38 0333 Be 7 0.01 B 4 750. C 4.8oo4 Fe 11 243 Cd. 7 2400. Xe135 4.3 3.5 x 106 U (nat) 8.2 7,42 3.92 u235 8,2 65o. 4590 u238 ~~~8,2 2.80 o. pU2 39 1025 664. a Total =a scatter + a absorption + a fission Absorption cross sections d~epend. upon neutron energy in two ways:. (I) 1/v d~epend~ence (2) resonance

Neutron Cross-Sections Cont. Examples of Velocity Dependence 105 ~4 104 Cadmium (Resonance) 103 '~ 2 tilo LA b 10 10 -._...\... 10'2 10l1 1 10 102 103 ev Macroscopic cross sections Z, L-NM Example calculation of N for carbon Atomic wt. carbon = 12.01 GM/GM mole Density = 1.65 Gm/ cc 1 GM mole contains 6.023 (1023) atomsN= 1.65 GM/CM3 (6.023) (10 23) atoms/ZGM mole 12.01 GM/GM mole 0.0827 (1024t) atoms/CM3 Calculation of Z. for Boron: N = pNa/M = 2.5 (6.023)(1023)/10.82= 0.2L39(1024) cm3 Za + Naa 0.139(1024) (750) (10-24) cm2 = 104 cnf' NeuronMen Fee at

Neutron flux = nv, s = scatter cross section -rsX 0 00 e SX number particles do= 0OES e-sX dx number scattered Average distance before collision x C00 xdo x0o Es e dx = x = _.... - - mean free path s r S -sx d0 0o zs e dx s 0o Xs average distance a single particle travels between collisions Example: Compute Xs for C = 48 (o10-24) CM2 N = 0.o827 (10)2 1 1 1 ES Nas.0827 (1024)(4.8)(10-24) 9~ Neutron Energy Loss on Collision Head-on Collision: V0 Before V.M V. After Glancing Collision.. V 0~~ 7711 Before After

i.e. Source Leakage + Absorption Now if q (E,r) is in introduced as the number of neutrons crossing energy E per unit volume per unit time then S (E) dE q (E + dE,r) q (E,r) (Er) dE an our balance equation becomes MD V2 + EaO _ q (Age equation in reactor theory) 6a Above the thermal energy region Ea _ 0 and also in this region as can be seen by the following argument. EE The number of collisions required to cross the energy interval E with average change in ln E of e is d ln E dE ~e Ee The number of collisions per unit time in this interval is E s (E) 0 (E) Thus, ZEOFE is the numaber that emerges from the interval per.unit time and. if there is no absorption, this quantity is equal to q the slowing daowm density. We have q -s Note that this gives a 1/B daependaence of flux on energy 10 OE ~~Thermal 1T

With the equation I (E) -- - our balance relation becomes cSE CEsE D 2 q -_ C~sE ~E And if the variables are changed so that dT - - CZs E We then obtain V2 q = 6q/6T [Fermi Age Equation no absorption]. Which, mathematically, is the same as the heat equation. If we have a point source of neutrons in an infinite medium, then by substitution q(rT) can be seen to obey the equation (r,T) = e -r2/4T [Slowi down density =________ from point source.. (47tT) 3/2 If the mean squared distance < r2(T)> 00 r2 q(T,r) dV < r2(T) > 0 00 q(T~r) dV 0 to reach a certain age 1- is computed,, it is found that < r2 (T-) > 6T or T is one sixth the mean square distance from birth at energy E0 to energy E. T~D dE. D dE CZ E ~EZ E Similarly if the balance relation for thermal neutrons (V2 6 7-u 0) is solved for a point source

becomes ~a 0 e 4t Dr and if < r2 >, the mean squared distance that a neutron travels from its source to absorption, it is found that 00 r2za r2 dr < r2 > 0 6ra.. Ea 0 r2dr 0 or if we call ED/Za the diffusion length then is one sixth the mean squared distance to capture. Scattering mean free path X, 1 s Z 5 of i s=n (Ef/IEt) (But Z5 varies with B) e: Z (avg) In spite of scatter neutrons advance from origin. The average cosine of~ angle 0 is a measure of advance. Thus,, we can define "transport mean free path! X 2 2 _S Xt+ C 35M3

td 2, Xt dE T=Lf Iz E J 3zf3s Et Average distance thermal neuts move before absorption is calle thermal diffusion lengtY = L /Xt a /1 at~ L = \13 \ t Ea 12. Neutron Shielding A reactor core during operation, Fission of U produces n averageof 2.5 neutrons/fission with the following spectrm S(E) dE - 2/,re SnM ',2E-e-E dE where (E) No neuts/fission neut in energy range fromn E to E + dE for high energy neuts (E 2 to 12 Mev). Above equation reduces to -0 3 72E S(E) dE = 33 e CIE 1, ci) ci) * 001-.0l l 10 E. Mev

Ideally, a reactor shield would be composed of those materials, in such amounts and distributions, that would reduce the levels of neutron flux, primary gamma flux, and secondary gamma flux to just less than safe tolerances at the outer reaches of the shield. Usually other considerations (in particular space, weight, cost, or structural strength) prohibit the use of such an optimum design. In such cases, one of these three fluxes will predominate. This particular one, say, for example, primary gamma flux, then becomes the limiting factor on the overall shield thickness; i.e., if the shield is thick enough to reduce the primary gamma flux sufficiently, all other radiations will be ore than adequately attenuated. Unfortunately, the discovery of any such possible predominance must generally await the performance of all or some of the calculations described below. 1. Neutron Attenuation.-The neutrons produced in a reactor are distributed in energy about a mean of approximately 2 mev; for shielding design purposes, their maximum energy can be considered as 14 mev. The neutron flux is reduced by the process of neutron capture, i.e., a neutron passing sufficiently close to a nucleus, forming a new (compound) nucleus. However, the probability of such a capture occurring (for most nuclei) is very small unless theenergy of the neutron is approximately thermal (i.e., a few hundredths of an electron volt). Thus, the majority of the neutrons must first lose energy before they can be captured. This energy degradation is accomplished by the process of scattering, both inelastic and elastic. A neutron cannot experience an inelastic collision unles-s its energy is greater than or equal to the energy corresponding to the first excited state of' the nucleus with-which it interacts. This threshold may be as low as 0.5 mev for elements of moderate or high mass numbers., but for light elements it is several mev. In fact,, hydrogen is incapable of causing inelastic scattering at any energy. ThereforeI in general, if' a scattering collision occurs below 1 mev i't will probably be elastic. On the other hand., inelastic scattering becomes increasingly probable as neutron energy increases. The process of elastic scattering of' high-energy neutrons does not contribute materially to the reduction of' the high-energy neutron flux. The energy lost by a neutron in an elastic collision is dependent upon the mass number of' the interacting nucleus and the angle through which the neutron is scattered. For a given scattering angleJ, the fraction of the neutron energy transferred to the nucleus increases as the mass number decreases. However, at high neutron energies the scattering cross section of an element of low mass number is quite small, since the scattering cross section at high_- _ enrg =__ 2nr- -> (2n) (1. x 2r _- 1013 x AI3) -1.4.5 1-2

Although at high neutron energies about one-half of the total scattering cross section of an element of high mass number represents elastic scattering, the scattering angle and loss in energy are both quite small. Hence, neutrons whose energies are in excess of 2 mev can best be degraded in energy by inelastic-scattering collision with nuclei of high mass number. A single such collision reduces the neutron energy to approximately 1 mev (the cross section being about one-half the total cross section of the heavy element). At neutron energies from 1 mev down to just above thermal, elastic scattering by hydrogen constitutes the most effective mechanism to attenuate neutrons. As noted above, inelastic-scattering cross sections are either zero or quite small in this energy range. Elastic scattering by hydrogen produces a greater degradation in neutron energy than by any other element. In addition, the scattering process is more nearly isotropic; this has the effect of lengthening the effective shield path of a given shield thickness. Thermal neutrons (those with energies of the order of magnitude of the average energy of the nuclei through which they are diffusing) are easily removed by capture in any of several elements which possess high thermal-capture cross sections. Of course, most of the elements in the shield can capture higher-energy neutrons; but, with the exception of certain narrow resonance regions, the corresponding cross sectionsare very low in comparison to thermal cross sections. Many of these capture processes are (n,y) reactions, some producing single gamma photons with energies in excess of 7 mev. This potential source of secondary gammas can be reduced by employing materials such as boron10. Boron-lO has a large (n a~) cross section. Although some of the liberated energy does appear as a gamma photon, its energy is only 0.5 mev. In summary, fast neutrons are best attenuated by materials of high density through inelastic scattering. Unfortunately., secondary gamma production accompanies this interaction so that materials for this purpose are placed close to the reactor core to allow shielding of these gamma rays. Usually one or more of the following elements are used for this purpose: tantalum, tungsten., thorium, lead., iron., or barium. Iron., having high structural strength, can be incorporated as part of the supporting structure of the shield. Where compactness is of chief concern., denser materials,. such as lead, are preferable. Intermediate-energy neutrons are most effectively attenuated by elastic scattering by light nuclei., which, for reasons of cost, limits the available material to one element, hydrogen (either as water, hydrogenous waxes, or plastic-s). Thermal neutrons require materials which have high thermal-neutron catrN — veh^-b- cros siect^+ i ons, svcho as4 boo-'lO,- lx.-n- and wic do% not yield h kard-cApturen i iY,4

As can be seen from the above discussion, the analysis of the neutron-attenuation processes is important not only fromn the standpoint of the reduction of neutron flux to safe levels but also as a means of estimating the distribution and strength of secondary gamma sources throughout the reactor, associated components, shielding, etc. 2. Gamma-Ray Attenuation. —Attenuation processes for gamma radiation have been discussed in detail in the previous section dealing with the gamma-radiation source. (See Chap. 4.) 13. Attenuation Calculations 1. Fast-Neutron Attenuation.-As can be seen from the above discussion, an exact analysis of fast-neutron attenuation in a reactor shield would entail a rather complex mathematical model requiring a prodigious amount of numerical computation. The present scarcity of experimental data has made it ossible to construct a theory containing only those elements which are physically most important. However, steps in this direction have been made, notably by Albert and Welton.* As previously noted, a considerable amount of hydrogen must be included in the shield to attenuate intermediate-energy neutrons. In a practical shield, collision with hydrogen usually has nearly the effect of absorption (insofar as required thickness is concerned). Qualitatively, this is true because of the degradation in energy which accompanies the collision, combined with the rapid increase of the hydrogen cross section as the neutron energy decreases. A small fraction of the initial collisions with hydrogen will give rise to neutrons having very nearly the source energy and almost their original directions-. These neutrons will cause the spatial distribution of neutron flux to depart slightly from the exponential form which would be valid if these first collisions with hydrogen were, in fact, captureprocesses. Likewise, high-energy neutrons which are inelastically scattered by an oxygen or a heavy nucleus may be assumed to be removed., since., as, noted pre - viously, its reduced energy will be something like 1 mev, the cross section of hydrogen at this energy being sufficiently high to insure rapid absorption. Similarly, at low energies isotropic elastic scattering is quite probable. Such collisions effectively change the direction of the scattered neutrons so that these neutrons give a small contribution at the outside of the shield. a. Effective Removal Cross Section. On the basis of the above argulments and certain exwperimental evidePnce,

heavier materials, it was found that the fast-neutron level generally decreases nearly exponentially through the shield. Thus, except for perturbations in the neighborhood of interfaces, the flux obeys an equation of the type neutron flux = constant x e-x; or, for a point source, constant -Zx flux = onstante x (2.13) 43r2 where x = shield thickness r = distance from source to point x, and Z = a constant. The constant,.Z Z R, is called the "effective removal cross section." Table 2.4 contains values of ZR for fission neutrons for a number of materials which have been measured recently at ORNL. An approximation of R/P = 0.085 A- '/3 where p is deilsity and A is atomic weight, is quite good for atomic weights above 10. TABLE 2.4. EFFECTIVE REMOVAL CROSS SECTIONS FOR FISSION NEUTRONS Cross section4 Cross sectionjt Material* barns/atom Material barns/molecule Aluminum 1.31 C7F16 26.3 (Boron) 0.97 C2F3Cl 6.6 Beryllium 1.07 CH2 2.8 Bismuth 3.49 B4C 4.3 Carbon 0.81 C30HG2 80.0 (Chlorine) 1.2 D02.8 Copper 2.04 (Fluorine) 1.29 Iron 1.98 'Lithium 1.01 Nickel 1.89 (oxygen) 0.99 Lead 3.5 Tungsten 2.5 *The effective removal cross sections for the materials in parentheses were derived from analysis of compounds containing these elements.

;The exact relationship for the attenuation of a beam of fission neutrons of unit 1source strength can be given in the following way* If S (E) dE gives the fraction of fission neutrons at E in range dE, then the number at a distance x from. a plane monodirectional fission source having 1 fission/cm2-sec penetrating through a mixture of water with other substances is 00 No (x) = ve ei x S ( (E) eH(E) XdE 0 where 4ri is the macroscopic removal cross section for the ith component of the mixture of water with other materials not containing hydrogen. or oxygen. These macroscopic cross sections are related to the microscopic cross sections by.602 GaH 60 62ipi ari ~~~~~~.602 0 orO 9 A.60 2 G arC 18 where the s ae in barns, G is the volume fraction of water in the mixture, the volume fraction of the other materials, pi and A their densities and atomic weights. An approximate solution to this equation can be developed by assuming that from 2 to 12 Mev aH is well approximated by aH= 5.13 'H (EQ.725) the fission spectrum byvs) 2 the integrand S(E) e -uHx at large distances is a sharply peaked. function of E which leads to the prediction that the maximumn of the integrand. occurs at an energy 0. 51.1 0Gx)58 Mev. with a full width at half maximum of A E. 1.55 (Gx)0'9Mv thus to caxry out the integration, the integrand. is represented. as a Gaussian., centered. at the peak energy 2 S (E) e-uHx = C (E-EO)

b. Saddle-Point Method. * For purposes of shielding calculations an approximate method (the so-called "saddle-point method") has been derived to estimate the attenuation f the fast-neutron flux by a thick-slab shield. Only the results of the derivation are presented here. The peak energy Eo of the neutron spectrum at a distance x centimeters from the source is Eo = o.541 (x)058 (2.1) where G is the volume fraction of water in the shield and E is energy (mev). The half-width, AE (spectrum width at one-half maximm value of the spectrum at x centimeters from the source, is given by AE = 1.55 (Gx)029 (2.1) Finally, the neutron flux, N(x), at x centimeters from an infiniteplane source emitting 2.5 neutrons/cm2-sec in a parallel beam i:s given by N(x) = 5.4 (gx)0'29e ox e-(1)rOx e-0.928 (X)58 (2.16) where F is the macroscopic oxygen-scattering cross section and Zr is the macroscopic scattering cross section of the heavy elements in the shield. While the above method is based upon a homogeneous distribution of materials in the shield., laminated shields can also be treated in this manner. In such cases the flux obtained at the outer edge of a given lamina is treated as a source at the inner face of the subsequent lamina. 2. Thermal -Neutron Flux Distribution. -As previously mentioned, a shield which adequately attenuates fast neutrons and gamma photons-will generally be more than sufficient as an absorber of thermal neutrons. However., capture gammas., activation, and heating effects may require an accounting of the thermal-neutron flux distribution. A definite phenomenological theory similar to the fast-removal theory is not available. Generally speaking., the thermal flux can be estimated by suitable bulk-shielding experiments. 5. Geometrical Effects in Fast-Neutron Attenuation.*Wigner, E. P., and Young, G.) Rept. MonP-283, Oak Ridge National Laboratory,

For a point source the geometrical factor 1/4rcr2 is all that is neede. a. Thermal Neutrons, Ideal Cases (1) For thick slab of shield - = 0 e-x/L where 0 = initial flux- neuts CM2-sec L = diffusion length, CM x = thickness of shield. CM Example 00 = 1010 neuts/CM2-sec shield = water L = 2.88 CM x = 2.88 CM = 1010 e-28.8/2,88 10 36 465 (110 05) 10 e./2.8 10 3.69 4i.65 (10o (a) Point source in infinite shield. =3no er/L e where: no= thermal. neutron emission =transport mean free path Note r in neutron. eq." is to first power rather than square as for gammas because of diffusion by scatter in the case of neutrons rat-her than straight line travel as for gamma Photons. 3 dr

(3) Point source in a finite sphere, radius R 3no (SINH (R-r)/L 4t Xtr SINH R/L no no -R/L R =4ARL (SINH R/L) 2 —RL 6 Example: Point source no = 10 neuts/sec R = 4 ft. = 61 CM shield - water find neutrons escaping n 4itR2j 2R eR/L 106 122 -61/2.88 ~ no ~ = 10 e 0.027 L 2.88 (4) Attenuation in material where a >> as (Example, cadmium and boron) 0 = 00 e-Kxwhere K = Ea C. Shielding Calculations for Thermal INeuts Using Neutron Leakage Factors take keff =l1= k T then (kcx T-l) =thermal neuts escaping core per k~ fast neut (1 -1) Lt leakage -K2L2 fission neutko WNo fPission -neuts+ PD(watts+) c' fissions neuts

Example Calculate leakage of thermal neuts from graphite-moderated thermal reactor: Power = 1 MW, k = 1.10, R = 250 CM, f = 0.9 2 ~ ~R2 K = (RA ) = (E)2 L2 = Lo2 (1-f) = 502(1-.9) = 250 K2L2 _2(250) 0.0362 neuts k-o- = (1. ) 250o2fi ssion neut No. fission neuts 106 watts (3 x 10 fissi watt-sec 16 neuts 7.5 (106) neuts sec total leakage = 0.0362 x 7.5 (1016) 15neuts = 2.7 (1015) neut sec d. Fast Neutron Shield~ing Calculations Using Leakage Factors Cross sections for fast neuts are small and are close to those pred~icted by the nuclear radius., r r = 1.4 (10-13) Al/3 CM Example: for lead A = 207 Approx. cross section = g r2 =t(l. 4)2 (lol13)2 (207 )2/3 CM =2.2 barns (measured value =3.5 barns)

2 2 no e- (r /4T ) (4(T) 3/2 where No. neuts that become thermal/sec in unit vol. at istance r from point source of no fast neuts T = Fermi Age or distance from fission to thermal energy oUsing the relation r2 q dv/ q dv Gives: r - 6 T 2 where r avg. square of r Neutron Slowing Down Distributions in Water and Graphite -o.16 0.1 ~~~~~~~~~2' Water T 3 3 0.1 -P 0.10 0.0 Ci) co 0.0 0

e. Use of Relaxation lengths in flux calculations for fast neutrons. - = Oo e X/ where X = effective relation length = 11 cm for ordinary concrete (p = 2.0-2.5) = 7 cm for hi-density concrete =3.5) = 4.2 cm for water Slope = -- 109 108 l~07~~~x 106 0~ 0) 5 10 C 1J 10.0 1023 10 1 *

After fast neutrons travel some distance from origin their attenuation approximates a straight line on semilog paper as shown on p. 7.25. The slope can be used to determine X, the relaxation length. Flux reductions for outer shields can then be estimated in a manner similar to use of 1/10 values in gamma shiedls. Example: Fast neutron flux = 1011/CM2-sec shield thickness = 6 ft. shield material = hi-density concrete, X = 7 CM x = 6 x 12 x 2.54 = 183 CM -X e/= 101 e (183/7) = 0 neut -x/ _ 1011 -(183/7) 0.5 CM2sec (tolerance 2(2 Mev) CM -sec Problem: A homogeneous reactor, 40 CM dia., has a fast neutron flux at the surface of 2.5 x 1011neuts/CM2-sec. A reflector of water 10 CM thickness is used around the spherical core followed by a shield of high-density concrete 150 CM thick. Consider reactor as a point source (producing 2.5 x 1011 flux at 20 CM radius) and calculate flux at outside of concrete shield. (Concrete contains 15 t. % H20, p.= 3.5.) 1. By relaxation length method = 4.2 CM for water reflector X = 7.0 CM for concrete shield 2. By saddle point method using ao for oxygen = 0.9 barns, ar for elements in shield = 2.0 barns; avg. atomic weight of heavy elements in shield = 120 0 29 -.928(Gx)0. 58 -Go x) = 4 Gx.. )_.. e j /r 1 r2 /r1 (For point source in spherical geometry emitting 2.5 neuts/sec.)

Solution to problem 1. -lO'4.2) -150/7 = 2.5 x 101l (e-10/42) (e ) =2 (ll) e 2-38 e-21.42 = 2 (11 2.5 (10 e) e-2'2 = 2.5 (10 )(4.13)(10 )(1.37)(lo (30/20 )2 (180/30 )2 = 0.141 neut/Cm2-sec 2. a. water attenuation =1 0.9 (10- 24)CM2 6.023 x 1023 atoms/CM mole (1 GM) atom 18 GM/GM mole (CM3) = 0.03 CM a = 5.4 (1)(10 29e 928(l)(10) 58 e(l)(0.03)(10) 00o ~ (3ocM/2o CM) 2 -= 5.4 (1.906) e-3'62 -0.3 = 4.57 (o.296)(.792) = 0.010 b. concrete attenuation = 3.5 (0.15) = 0.525, G-1 =.475 Z, = 0.03 CM ' Kr- 2.0 (3.5) (6.023-)(lol1) = 0.035CM 120 __b 5.4 (.525) (lo)~ '29 e 928 (.525 x l50) -58) ~525(.03)(150) -.475(.035)(l50) 22 ~~( )e')(e) 00 1802 /302 0.33 (.83 x10 %)(.095) (0.9827) =2.09 x 1 -8 =2.5 (loll1)(0.0l)(2.09)(10 8) 52 neuts/CM 2-sec (Note: Increasing water content of concrete reduces flux.)

DOSE ltRc di R 4 POINT REACTOR CORE IRON WATER Geometry of spherical core and shield.* If the core and shield are shown with a uniform power distribution (p watts/cm3) in the core, the uncollided fast flux at distance R centimeters beyond the surface of the core is ~~~~~~~ 1Al~ a~a+ a Jibc =U(R)= 2~x(7o75)(1lO)10 P) 1 R/ d.iP ab p2dPG(Zcpc)G(Zsps), (2.17) where 7.75 x 101~P(watts) = fission neutrons/cm3-sec, a = R +Rc, bc = a2 - Rc, I = Cos 9 G(p) = the point-source attenuation kernel, G(Z pc) = the point-source attenuation kernel for the core, and G(,Sps) = the point-source attenuation kernel for the shield. There are two possible choices for G(Zp): Equation 2.14or an exponential fit to an experimentally determined distribution. The latter will have the form N G(p) = Z Ai ezip ~ (2.18) 4tp2 i=l If the volume fractions of materials other than water are fc and fs in the core and the-shield, respectively, Equation 2.19 reduces to * Rockwell, Theodore, III, "Reactor Shielding Design Manual", TID-7004, March 1956 7.7.28

p2N __ _ __Ai.875(10)1~ 2P RC 1=1 fc(Nczi)+i 2 C i =1 f c (Zinc - 'i~ ) + F'i exp {- [fs(ms -Zis) + Zis] RJ - exp {-2[fc(ZIc -Zic) + Zic] RC) (2.19) where = Ps the distance from unit-source volume to surface of sphere along line between unit source and the point R Ps. pa;2f2-bc = the distance in shield along line between unitsource volume and point R, c, = the effective removal cross section for the nonhydrogenous materials in the core (cm-'), and = similarly, for the shield. b. Cylindrical Core and Shield. * If cylindrical coordinates are used (see Fig. 2.4), the ucollided flux at distance R centimeters beyond the surface of the core is r~~ pc ' s(r, z) rd~rdrdz G(R),(2.20) 2ic J1=o dr=o dzp2 where 2 2 2 2 p =z + a + r -2ar cosAF = the azimuthal angle with respect to the transverse axis of the cylinder, and a = the distance from the axis of the cylinder to the point at which the flux is being calculated. *Rockwell, Theodore., III, "Reactor Shielding Design Manual, TID-70-,Mrh15 7.7.29 ~ ~ ~ -00yMac 15

7.8 Experiment No. 8 Germination and Growth of Irradiated Seeds and Sprout Inhibition in Tubers Discussion Authorities differ widely in their opinion onthe use of ionizing radiation to affect the germination and growth of plant seeds. In an article by Kuzin (1) four different methods of irradiating treatments were tried with varying degrees of success. 1. irradiation before sowing 2, soaking of seeds in solutions containing natural and artificial radioactive substances 3. treatment of the soil with radioactive substances serving asmico fertilizers 4. continuous irradiation of growing crops withraiation Kuzin claims that irradiation of seeds prior to sowing offers the greatest advantages as compared to the other methods of treatment. Its chief merits are1. Irradiation can be performed in specially equipped places 2. Irradiation of the seeds can be completed in a selected finite time. 3. Complete absence of radioactivity both in the sowing material and in the yeild., Also, the acceleration of the initial stages of germination Is important since it may influence the yield in arid districts as well as those where the sowing period is limited. Observations at the University of Michigan have indicated some of the effects of radiation on seeds and tubers (see Chap. 9 of "Radiation-Uses in Industry and Science")., In studies in production of mutants, seed walnuts were irradiated at various dosages from 0 to 50,000 rad, labeled,, placed on a mat of sawdust and covered with several layers of damp burlap to hasten germination. Inspecti~ons were made twice weekly to check germination progress and to dam-~ pen the walnuts and burlap. The following table summarizes the early observations on germination of walnuts.

Observations on the Germination of Irradiated Walnuts Number Dosage, Germinations, Germinations Number growing, stratified, Krad May 24 total, Aug. 8 Aug. 22 Mar, 27 July 3 50 0 3 7 5 5 50 1 2 4 4 3 50 5 3 6 6 6 50 10 2 8 7 7 50 20 2 10 9 9 50 30 2 16 12 10 50 40 3 16 3 1 50 50 3 18 1 0 In student experiments at the Fission Products Laboratory radish seeds were irradiated at various doses and planted in a small plot of ground outSi'de the laboratory. Figure 7.24 shows the experimental radish bed. -Reading from right to left the radiation dosages were 0, 400 rad, 1,000 rs~d, 5,000 rad, 10,,000 rad, 50,,000 rad. Typical plant specimens were removed when the plants were f our weeks old and are shown in Fig.- 7.25. These ex.periments were inconclusive because of lack of suf ficient control of the variables, but limited dat-a indicated that a dose of 1,000 rad appeared to produce the most vigorous plants. These and other observations indicated a two-f old effect of ra iation~ (1) 'Interference with cell division at doses of 5000 rad and greater., as 'indicated by inhibition of continued growth of sprouts; (2) stimulation of growth hormones as shown by more rapid sprouting of irradiated onions and hormone studies on potatoes. However, 'if the radiation dose 'is kept suff iciently low, 'it is believed that stimulation of growth may be effected without significant interference with cell division. In the Russian studies the most benef icial dosages usually ranged from 300 to 1000 roentgens.

observation is theoretically important because the increase in root diameter is due not to the growing size of the cells, but to the increase 7 -8.- 2a

In student experiments at the Fission Products Laboratory radish seeds were irradiated at various doses and planted in a small plot of ground outside the laboratory. Figure 7.24 shows the experimental radish bed. Reading from right to left the radiation dosages were 0, 4 rd, 1,000 rad, 5,000 rad, 10000 rad, 50,000 rad. Typical plant specimens were removed when the plants were four weeks old and are shown in Fig. 7.25. These experiments were inconclusive because of lack of sufficient control of the variables, but limited data indicated that a dose of 1,000 rad appeared to produce the most vigorous plants. These and other observations indicated a two-fold effect of radiation: (1) interference with cell division at doses of 5000 rad and greater, as indicated by inhibition of continued growth of sprouts; (2) stimulation of growth hormones as shown by more rapid sprouting of irradiated onions and hormone studies on potatoes. However, if the radiation dose is kept sufficiently low, it is believed that stimulation of growth may be effected without significant interference with cell division. In the Russian studies the most beneficial dosages usually ranged from 300 to 1000 roentgens. Rye seeds given a dose of 750 rad, produced roots having a diameter of 393 ) on the 3rd day of development as compared to 304 for nonirradiated rye seeds used for control purposes. This observation is theoretically important because the increase in root diameter is due not to the growing size of the cells, but to the increase in their number. For example, the number of cells in the subepidermal layer of shoot roots was 68 for the rye seeds given 750 -rad dose as compared to 4o for the control rye seeds given 0-rad dose. The root length on the 4th day of development was 52.2 mm for the 750-rad rye seeds as compared to 43.5 mm for the control. Similar results on stimulation of plant growth have been observed with other species. On the 5th day of development radish, pea, and cucumber shoots from seeds given an X-ray dose of 500 rad had shoot length of 72.2, 61.2, and 89.5 mm., respectively; whereas the respective untreated seeds had shoot lengths of 50.0, 53.6, and 82.6 mm, respectively. Yields per unit field area were also increased. In the case of radishes grown in a greenhouse a dosage of 1000 rad gave a 30 per cent increase in total weight of tuber yield as compared to the controls. Field grown radishes from seeds receiving the same dosage gave a 4o per cent increase as a result of irradiation. Cabbage seeds given a dosage of 1000 rad produced plants that ripened earlier then the control and gave a 19 per cent increase in yield per area. The harvest yield of peas given an X-radiation dose of 350 rad produced 10 per cent more pea seeds per plant. Furthermore, the weight of 1000 pea seeds was 16 per cent greater for the yield from irradiated plants than from the controls.

harvest grains. Low doses of ionizing radiation (5,000 to 20,000 rad)have been used to prevent the sprouting of tubers and bulbs during storage. Figures 7.26 and 7.27 show irradiated and nonirradiated potatoes and onions respectively stored for a few months after irradiation. Onions and potatoes seem to respond differently to irradiation. During initial storage both irradiated and control onions developed small sprouts. In the case of the White Pearl onions the irradiated onions sprouted first. In all cases sprouts on irradiated onions grew only a short distance (about 1/2 to 1 1/2 in.) and then withered and died. On the control onions, sprouts continued to grow up to 6 to 8 in. All irradiated onions remained dry and firm (but lost some crispness) whether or not they had sprouted; control onions became moist and soft. The contrast in appearance of the control and irradiated Bermuda onions is shown in Fig. 7027. This photograph shows the long sprouts of the control onions protruding several inches and no sprouts on the irradiated onions. Tests of the irradiated onions after storage indicate some loss in crispness and some loss in pungency. Brownell et al. concluded that a 7-krep dosage of gamma radiation may not completely prevent Initial srouting but it inhibits the continued growth of sprouts, keeping the onion more firm and increasing storage life. Studies of Potato Irradiation 'in Europe Studies on irradiated. potatoes heave been underway in Russia, since 1955. Extensive investigations have 'been made on the chemistry,~ odor., taste, and cooking qualities., On the basis of these tests the USSR Chief of Public. Health Inspection authorized the use of potatoes given a, dose of 102000 rads for human consumption. A large capacity radiation facility has been, constructed, and the process 'is in use. Norwegian studies indicated, that radiation doses as high as 1.6)000 rads could be used on potatoes. Similar studies 'in France indicated that doses higher than 10,000 rads 'increase spoilage and that doses of about 7500 rads are.to be preferred. In 1959 Vidal of France commented that "For short term storage doses of 5000 to 6000 rads are sufficient but for long term storage (12 months or longer) it is necessary to use doses of 7500 to 10,000 rads." Burton and de Jong of England reported extensive studies on irradiated Ware varie+ t Y v% + ptoes. r Tkhey st telmhe-ffec of- i-rrtAadiation ( h ig clevel-,r I is cthusn1

7.24 Plot of growing radishes from irradiated seeds (reading right to left, control, 400, 1000, 5,000, 10,000 and 50,000 rad). 7.5Typical radish plants from irradiated seeds. (Top row left to right, control, 400, 500, 1,000 rad; bottom row 10,000 and 50,000 rad.) 7.8. 4a

? -i;.!;:: 7' 111:1..V i;;'AM - 0 S AS -i:-:i: eV a *% i;; i 0 0....:*:. di: ~:..-. i t:.,- i NON-IRRADIATED 21t,000 REP Fig. 7.26 Irradiated and nonirradiated potatoes (irradiated in May and stored four months before photographing). i.4M, NOT DECAYED Fig. 7.27 Irradiated and nonirradiated onions (irradiated in May and stored four months before photographing). 7.8.4b

mercially acceptable sprout suppression —certainly less than 10,000 rads." American studies have shown that the permissible dose varies ap;preciably with variety, length of storage before irradiation, and storage conditions. This may explain some of the differences in doses recommended by European investigators. Procedure a. Sprouting inhibition Immediately after harvest tubers, bulbs, and other root crops have a period of dormancy during which sprouting does not occur whether irradiated or not. Therefore,in demonstrating sprout inhibition by irradiation, root crops should be used that have been in storage for a sufficient period of time for sprouting to be imminent. Autumn harvested potatoes stored until spring and onions stored for about three months are usually satisfactoryo Sweet potatoes, carrots, beets, and turnips showing small sprouts may be usedo If the tubers have small sprouts, the buds or sprouts may be brushed off so that none remain. The tubers may be divided into a number of similar groups and radiation doses given each group in increments of about 5,000 rads. After irradiation, the tubers should be stored and observed for evidence of sprouting and/or rotting. At the first sign of development of rot the tuber involved should be disposed of to avoid spread of spoilage organismso Onions will keep best when stored at temperatures not below 50 degrees F. and humidities not above 70%. Potatoes store well at refrigerator temperatures of cabout 35 to 40 degrees F. and humidities of 90%. b. Seed germination Most workers in radiation biology agree that limited doses of radiation are beneficial in stimulation of the germination of seeds particularly seeds that sometimes show poor germinatioon. For experimental purposes a variety of seeds may be selected and irradiated at doses in the range used by Kutzin (200-29000 rad)o A few larger doses (10,000, 50,000, 100,000 rad) may be used to demonstrate harmful effects of excessive irradi.ation. Extensive data on germination alone may be obtained by placing small seeds between dampened pieces of blotting paper. The larger seeds and nuts may be placed in dampened. sawdust, peat moss, or vermiculite. The seeds shoul.d be kept moist until germination is complete. Germination tests alone do not indicate whether or not better growth and better yields are possible as evidenced by the early death of the irradiated walnut seeds receiving 40 and [ krad doses (see table). These seeds germinated well but died shortly after germination. To obtain data on growth and yield the irradiated and control seeds must be planted in suitable containers for hot house studies or in suitable plots of ground outsideo Many variables are involved which has resulted in wide differences in the reports of various investigators as to the benefits of an irradiation treatment on crop yields. 7..;. 5

7.9 Experiment No. 9 Food Irradiation Discussion 1. Radiosterilization —Many studies have been conducted in various laboratories using various doses of ionizing radiation for food preservation. The initial studies were in most cases directed toward the use of "sterilizing" doses in the range of 2-5 megarad. Preliminary studies indicated that sterilizing doses of radiation produce undesirable changes in most foods, particularly in the odor and flavor of meats and dairy products. Many investigators in various laboratories are studying the problem and satisfactory solutions have been found for certain food items. Radiosterilization which permits long-term storage at room temperature is the ultimate goal of much research being conducted in this field. However, it presents more problems than the use of lesser dosages of radiation which would not completely sterilize. 2. High-Radiopasteurization —Limited studies have been made at the University of Michigan using a "high-radiopasteurization" dose of gamma radiation of about 1 megarad. This dose is not sufficient to sterilize completely, but is sufficient to preserve food. Thus, this process could not be used where the possibility of growth of Clostridium botulinum exists. Destruction of the spores of this organism is the first requisite in successful thermal processing of canned foods. However, if the danger of botulism is avoided by control of pH or oxygen level and temperature, a dose of 1 megarad would be sufficient to preserve food from spoilage by vegetative microorganisms, mold, yeasts, etc. In limited tests, smoked fish (salmon and chub) packaged in 2-mil heatsealed polyethylene bags and given a radiation dose of 1 and 0.8 megarad, respectively, "kept" at room temperature for several months. The irradiated salmon and chub were considered to have very good flavor immediately after irradiation and the flavor was still, good after storage for one month at room temperature. After 3 -mo. storage at room temperature, the salmon developed a slightly rancid taste. Packages of salmon that received l-megarad dosages and eventually spoiled did so as a result of mold growth. Fig. 7.28 shows a photograph of the irradiated salmon and control after storage for nine days at room temperature. The samples of salmon were removed from the polyethylene bags for photographying. Note the development of mold on the control sample. Mold of this type first appeared after storage for about three days, whereas no mold growth appeared on the irradiated salmon. Fig. 7.29 shows a plot of the original salmon samples and, although only a few samples were used in this limited study, the curves are indicative of subsequent data obtained with larger numbers of samples 7.9 ol

Fig. 7.28 Sample of Irradiated and Nonirradiated Smoked Salmon after Removal from Sealed Polyethylene Bags. i,Ooo,ooo REP 100 CONTROL \ \10,000 REP 2/4 2/8 2/12 2/16 0 4 8 DAYS 12 14 16 Fig. 7.29 Irradiated smoked salmon data. Salmon was stored at room temperature. 7.9.2

3. Radiopasteurization ---A dose of radiation less than that required for sterilization and in the range of about 100, 000 rad might be termed a radiopasteurization dose in that, like thermopasteurization, most but not all of the microorganisms are destroyed. Radiopasteurizating doses equal to 2-5% of sterilizing doses can be used to reduce the population of microorganisms 90-99% without producing undesirable changes in the food. Such a radiopasteurization treatment could be used to extend the refrigerator shelf-life of perishable foods which might make possible new methods of handling. A new method of wholesaling fresh meat has been proposed for consideration by some of the larger packing houses and some of the larger retailers of fresh.meat. This proposed new method consists of preparing packaged standard cuts of fresh meat, packaged fresh ground meat, packages of cut chicken, etc., in retail-size portions at the packing house rather than at the retail meat market, and. of radiopasteurizing the packaged meat at the packing house by using a dose of about 100,000 rad prior to shipping to the retailer. The extension of the refrigerator shelflife of fresh meat by radiopasteurization should make this new method feasible. A number of advantages might be realized by the consumer, the retailer, and the packing house as a result of using this new method of wholesaling meat and radiopasteurization. Since microorganisms play a major role in the spoilage of meat and meat products, their storage life at refrigerator temperatures should be increased by keeping the microbial flora at a minimum* Rigid satitation practices by meat packers and meat processing plants have done much to increase the storage life of meats, but the subsequent exposure of meat to air and contaminated surfaces makes the control of the microbial population a difficult problem. Radiation dosages needed to inactivate vegetative forms of microorganisms are much lower than those required for bacterial spores. Decreasing the vegetative microbial flora sufficiently to prolong the storage life of meat at refrigerator temperatures should require a much smaller dose of gamma radiation than is required for sterilization and destruction of spores, In some tests at the University of Michigan grams of fresh lean ground steak were aseptically transferred to flasks containring sterile glass beads and sand. Prior to irradiation the meat was inoculaeted with 1 ml of a suspension of a psychrophilic gram-positive bacterium originally isolated from spoiled ground beef. This organism grew readily at a temperature of 4~ C. An uninoculated control sample was kept to establish the meat's normal population Table 1 shows the microbial counts on the meat immediately after irradiation. No significant differences in counts are apparent up to a dosage of 40,000 rep. Table II shows counts per gram of meat after subsequent storage at 4~ C. The data are plotted on semi-logarithmic paper in Fig. 7.30 as total counts per gram. 7,9.3

lo, - - 0-) -- - 0! 1,' 0 -- -. [. A. CONTROL INOCULATED 0 20,000 REP 0 40,000 REP -,. A 80,000 REP - - -~ _ 3 160,000 REP 101 O 1 2 3 4 6 7 8 9 10 11 12 13 Days storage at 4 degrees C Fig. 7.30 Growth of microorganisms in irradiated meat,I/ I / o.......!........... Fig. 7.31 Irradiated (left) and nonirradiated (right) grapefruit(previously infest with egge of the Mexican fruit fly)after development of mature larval stage. Note secondary damage from mold in the control fruit. 7.9.3a

TABLE I-EFFECT OF GAMMA RADIATION ON AEROBIC MICROBIAL FLORA OR GROUND BEEF INOCULATED WITH 3.2 x 104 PSYCHROPHILIC BACTERIA PER GRAM Microorganisms Sample Per Gram % Survivors Control 8.0 x 1O4 100.0 20,000 rep 1.4 x 105 100.0 40,000 rep 6.8 x 104 85.0 80,000 rep 7.5 x 103 9.0 160,000 rep 9.5 x 103 12.0 TABLE II-EFFECT OF SUBLETHAL DOSES OF GAMMA RADIATION ON THE AEROBIC MICROFLORA OF GROUND BEEF AFTER STORAGE AT 40 C Number of Microorganisms per Gram of Meat After Sample Indicated Number of Days of Storage at 40 C 0 2 4 8 13 Control 5.5 x 104 8.0 x 106 5.5 x 108 3.3 x 1010 2.8 x 1010 Coatrol Inoculated 8.0 x 104 x106 4.3 x 107 2.5 x1010 4~5 x 1010 20,000 rep 1o3 x 105 2.2 x 105 5.0 x 107 --- 2.8 x 1010 40,000 rep 6.8 x 104 6.5 x 104 1.8 x 106 3.0 x 109 8.3 x 109 80,000 rep 7.5 x 103 2.0 x 104 6.3 x 103 9.3 x 106 6.5 x 109 160,000 rep 9.5 x 103 2.8 x 103 4.3 x 103 1.2 x 106 13 x 10 7.9.4

No detectable differences in either color or odor were noted in the samples immediately after irradication. After five days of storage, both the inoculated and uninoculated controls had a pronounced putrid. odor. Initication of spoilage in the sample receiving a radiation does of 20,000 rep was indicated by a slight off-odor. After 13 days of storage, all the irradiated samples except that which received a dose of only 20,000 rep were still free of off-odors and there was no indication of spoilage. 4. Low Radiopasteurization —A still lesser dosage of radiation or 10,000 -25,000 rad termed low radiopasteurization might be used for some other types of food processing. Such low-radiopasteurization doses will not destroy a sufficient quantity of the population of vegetative microorganisms found in fresh foods, such as meat, to affect appreciably the storage life at either refrigerator or room temperatures. This range of radiation, however, has the least influence on flavor and color and would be the most economical to employ because of the high radiation capacities possible at such low dosages. Such a dosage might be used for the treatment of potatoes and onions for sprout inhibition, for grain, cereal products, and fruit (See Fig. 7.31) for insect control, to break the cycle of trichinosis in pork, to control tapeworm from beef, pork, and fish, and to control a number of other diseases caused by parasites. The irradiation of potatoes and wheat, as well as other grains, is considered to be feasible on a mass scale from the viewpoint of both engineering and economics. Because of the tremendous quantities of these staple foods, the application of radiation would be on a large scale. Procedure The experimental demonstration in a precise manner of the preservative effect of radiation on various foods is complex since it involves the determination of populations of microorganisms, molds, and yeasts, that might cause food spoilage. Such studies can only be performed with precision in a laboratory equipped for studies in bacteriology. Also, the evaluation of the quality of irradiated food is best investigated by the use of skilled pannels and a statistical number of evaluations. However, for purposes of demonstration a number of simple experiments may be performed. Dairy products and meats are particularly susceptible to flavor, odor, and/or color changes as a result of irradiation. In one experimental demonstration of flavor changes induced by irradiation replicate samples of pasteurized milk, American process cheese, butter and a cooked meat product may be given radiation doses of 0, 50,000, 100,000, 200,000, and 400,000 rads. After irradiation the flavor, odor, and appearance of the samples receiving the various d.os ages should. be compared t1o thle control s ample (0 dosage;). Most individuas report a threshold for off.-flavor in irradiated milk at about 50,000 rad.s. Irradiated butter and other fats hIave a similar off-flavor threshold, but the rancid flavor of irradiated fats may be removed in many cases by heating 7.C).5

the fat after irradiation. Recooking meat after irradiation also is beneficial to the flavor of this product. To demonstrate the preservative effect of radiation from spoilage by microorganisms fresh bread slices may be packaged in polyethylene bags, heat sealed and given radiation doses of 0, 200,000, 400,000 and 800,000 rads and stored for several weeks at room temperature. The control (0 dose) usually molds in about one week at room temperatureo The sample receiving 200,000 rad may not show mold growth for several weeks. Samples receiving 400,000 or higher doses may not mold during storage for months. Fresh strawberries and lemons may be used to show the use of radiation for the prevention of mold growth on the surface of the fruit. In this experiment, boxes of strawberries may be given 0, 50,000, 100,000, and 200,000 rad doses of radiation and then stored in the refrigerator. After one week or less of storage, the nonirradiated strawberries may mold, whereas the irradiated fruit may be stored two to three times as long before mold appears. Similar observations can be expected with treated lemons stored at room temperature except that the time for mold development may be longer. The unspoiled irradiated bread and fruit also may be tasted for evidence of flavor change. 7 9 o6

7.10 Experiment No. 10 Effects of Radiation on Chemical Reactions by N. C. Kothary The use of electromagnetic radiations to induce and accelerate chemical reactions has been known for several decades. However, in the past this phenomenon was limited to the use of ultraviolet light which was recognized as the chief component of sunlight effective in promotion of chemical reactions. The rapid growth of the use of radiosotopes and nuclear radiations has stimulated interest in the effect of other types of radiation on chemical reactions. There are several advantages associated with use of radiation for promotion of chemical reactions. Certain reactions can be induced which are not possible otherwise. Certain other reactions normally require a very high temperature or catalyst which could be eliminated or the requirements could be reduced to more convenient values by use of radiation. Discussion In the present experiment the study is limited to use of cobalt-60 gamma radiation in synthesis of polysulfones. Radiation Chemistry: Gamma radiation in the energy range used does not induce radioactivity. The major primary effect is the liberation of a number of electrons as a result of absorption of radiation by matter. These electrons may lose their energy by exciting or ionizing molecules along their paths. The subsequent roles played by the excited and ionized molecules in producing the final chemical change is not understood completely. In many chemical systems however, it is shown quite conclusively that the reactions are typical of free radicals which are uncharged molecules with unpaired electrons. These radicals are produced by the eventual decomposition of excited or ionized molecules and free electrons, and are usually extremely reactive. Certain chemical reactions are of the chain type and suffer a loss in free energy due to the chemical change. In such a case the initiation of a single chemical reaction by radiation may result in cleavage and formation of a large number of chemical bonds. This results in a highly efficient use of radiation energy. This is an important factor because of the. high cost of this form of energy. Radiation induced polymerization is a chain reaction giving high yields and the products obtained may show high purity because of the absence of catalyst. There are certain peculiarities of radiation induced chemical reactions which proceed by radical mechanisms: 1. The reaction is sensitive to the presence of oxygen which reacts with the free radicals. 7.10.1

2. The reaction is also sensitive to slight amounts of impurities which may act as radical scavengers. 3. Reactions in the condensed phase will differ from those in gaseous phaseo An important effect known as Frank-Rabinowitch "cage" effect states that if two entities are formed due to radiation and are unequal in size, the larger fragment has a smaller probability of leaving the "cage" of surrounding molecules, before reacting furthero 4. Another effect known as "sponge effect" explains the greater stability of benzene or other aromatic structures as due to the larger number of energy levels of the phenyl groupo The absorbed energy could be dissipated throughout the molecule in such a case, rather than cause a chemical change. 5. The rate of polymerization is most often proportional to the square root of the intensity (I0o5) of radiation. This dependence is typical of many radical polymerizations where the termination step occurs by the combination or disproportionation of two growing chainso In most of the work the yield of a chemical reaction initiated by radiation is usually expressed by what i~ known as G value. This is defined as the number of molecules converted per 100 eV absorbed, Synthesis of Polysulfones Braylhas studied the copolymerization of sulfLr dioxide with olefins. The resulting polysulfone is a thermoplastic resin having many physical properties similar to those of some commercial plastics. He has suggested the following steps for the reaction. 1. Production of Compton electrons. 2. Ionization and excitation of molecules by dmpton electrons. 3. Neutralization of ionized molecules by thermalized electrons producing excited molecules0 40 Decomposition of excited molecules to free radicals. 5. PropAgation and termination of free radicals in the copolymerization reactions. Barb has studied the copolymerization of styrene with sulfor dioxide initiated by ultraviolet light. He found that in general the polymer was composed of two molecules of styrene to one of sulf4r dioxide and the following 711 2

structure is indicated. Yemin at the University of Michigan is now studying this reaction using gamma radiation instead of ultraviolet light. The copolymer formula is: 0 CH - CH - CH - CH - S C6H5 C6H5 Procedure (1) Purification of Reactants As mentioned in the discussion, and indicated for the experiment on dosimetry using ferrous sulfate, small amounts of impurities in the reaction system are not desirable. Hence, the reactants used should be of reagent grade and purified further if possible and found necessary. The styrene used for the reaction is reagent grade styrene supplied by Eastman-Kodak Company and purified further by distillation in a nitrogen atmosphere under reduced pressure. This is also necessary as commercially available styrene contains an inhibitor which is added to prevent polymerization under normal storage conditions. The sulfur dioxide used is Matheson reagent grade and is purified further during the loading procedure used. (2) Loading of Reactants A heavy walled pyrex vial is used as the reaction vessel. It is constructed from heavy walled 14 mm pyrex tubing with one end sealed and the other drawn out to a constricted neck and 7 mm open tube. The use of heavy walled tube is necessary as the reactants develop super-atmospheric pressure at room temperature. (See Fig. 7.32o) The vials are cleaned with chromic acid cleaning solution and rinsed with distilled water and reagent grade acetone and finally dried at 120 degrees C. After weighing the vial, a known volume of styrene is introduced by siphoning from the distillate receiver. The vial is weighed again to find out the exact weight of styrene used. It is then attached to the vacuum system. (See Fig. 7.33.) The vacuum system is a glass manifold with inlet valves for reactants, manometer and the vacuum pump, The whole system is evacuated for some time after immersing the vial with styrene in dry ice. At this temperature (-79 degrees C.) the vapor pressure of styrene is negligible and thus, negligible losses occur. After evacuating the system of residual gases to a pressure of 10-1 mm, the sulfur dioxide valve is opened and the reactant is condensed in the vial. Again the reactant is measured by volume and weighed l ater to check the exact amount. 7.10.3

o.~ ti09 0" O -4(h> ~ ~ ~ ~ ~ ~ ~ ~ ~ L Br~,8mm.STD.WALL (-'C~ ~ ~~~~~~~~~~3 r\)IU~ PYREX TUBE - > z -'NO H' H' NO r 'P r NECK, HEAVY WALL SPECIFICATIONS Oq C tl (T9 ~ -10 mm. C. D. Gq F. A (Jtj3 mm. I.D. A 3 - "3 mm'.I.D C 0 SEALED WITH HEAVY 0 FE) WALLED TIP M P ci- CD < vlSHOULDER CONSY. CD C~C) H'~ H1) ED H' N 4 H CD~C - 0 o H~~~~~~~~~~~ m A~~~~ -i -+ (D a ()0 COC ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~0 f-0.D ~-mCI~~~H WAL L c+ (D HF (D _ _ _ _ _ _ _ _ _ _ _ _C H'n~~~~~~~~~~~~ 1r~~~~ N Co CD~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~( Ce < a 0 M L-zj Z 3 0 B- e ASE ROUNDED ~3D CONST. mTCK. HEAVY WALL U1~~~~~~~~~~~~~2 r -/ 53 -/ (D (D CD -4 -Ft P(D r Reactor 0* Do Wo T. L, arr ao r,, Co 0 ' Designation Inches -Inches Inchei 0 ~ ~ ~ ~ ~ ~ ~~~~~~~~~~~0 b n ct-~~~~~~~~~~~~~~~~~~~C A34 1/8 3-3/4 0 B 1-1/2 5 /32 5-1/2 ct cn~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ 5]/32 6 (D~~~~~t (D i CO (D et (D ~ ~ ~ ~ ~

After the loading is completed, the vial is sealed off at the neck with a torch. The open tube end of the vial is saved for subsequent weight measurements. (3) Irradiation of Vials The vial is placed in a Dewar flask containing a constant temperature bath (0 degrees C. provided by ice in water) in the center well for 6 hours. The radiation intensity and total dose should be noted from previous experiments. (4) Recovery of Product After irradiation, the vial is cooled and opened by scratching the vial near the neck and touching one end of the scratch with the molten end of a pyrex glass rod. It should be noted that improper opening of the vial may result in shattering and explosion of tube, and release of poisonous sulfur dioxide. The whole operation should be conducted in a fume hood. The contents of the vials are allowed to thaw and are poured.while still cold into 200 cc of methanol containing hydroquinone. Methanol is used as solvent for the monomer. The polymer formed due to reaction is not soluble and precipitates out. The hydroquinone (a free radical scavenger) is added to stop any reaction which might proceed as a post irradiation effect.The solution is filtered through a Buchner funnel and precipite recovered on a preweighed Whatman #42 filter. The filter and the precipitate are dried and weighed, to determine the amount of polymer formed. Questions 1. Determine the percent yield of the reaction and the "G" value for the same. 2. Briefly discuss the significance of various factors that might be considered in determining the economic feasibility of a chemical reaction promoted by radiation. References 1. Bray, B.G. "The Effect of Gamma Radiation in Several Polysulfone Reaction" Ph. D. Dissertation, University of Michigan, Ann Arbor, 1957. 2. Barb, W.C., Proc. Roy. Soc. London, A-212: 177092 (1952) 3. Kothary, N.Co, and Yemin, L. "The Effect of Gamma Radiation on Several Reactions of Unsaturated Organic Compounds", Engin. Research Institute, University of Michigan, Ann Arbor, 1957 7.10.5

7.11 Experiment Noo 11 Glass Dosimetry-Cobalt Type Discussion Ferrous-ferric (5-50 Krad) and the cupric-cuprous ( 10 Mrad) dosimetry require a certain degree of operator technique to produce consistent results. There are inherent problems assiciated with any form of liquid dosimeter that are avoided by the use of solids for dosimetry such as cobalt glass (Bausch & Lomb F - 0621. This method of dosimetry provides a relatively simple and positive means of measuring doses in the region of 104 to 106 rad. Under irradiation the cobalt glass and other dosimetry glasses such as silver activated phosphate glass develop color centers which result in an increase in the absorption spectrum between 3)000 A and 59000 Ao However, the optical density of the silver activated phosphate glass is less stable with them than the cobalt glass. At high doses there is some fading and also a noticeable falloff in the curve of dose vs absorption coefficient. This indicates there is a saturation of the color centers occurring. Higher temperatures will hasten fading of the high dose sampleso (See Chapter 3 for additional information.) Figure 7~32 shows a plot of optical density vs dose for Bausch & Lomb cobalt glass. Table 7. lists Bausch and Lomb specifications and prices for cobalt glass, silver phosphate glass and accessories Table 70 Specifications for Bausch & Lomb Cobalt Glass and Silver Activated Phosphlate Glass and Accessories Catalog No. Description Price Each 33-99 Cobalt dosimeter glass) 15mm x 6mm x 1.5mm, faces polished, edges rough grind... $.30 33-99 Silver-activated phosphate glass, 15mm x 6mm x 1o5mm, faces polished) edges rough grind....40 33-66-02 B & L Microdosimeter Reader 1!0-10,000 rads, for 60 cycle) 155 volt A.C., including directions for use) plastic dust cover) and one pair Microdosimeter Rod tweezers,., 1,490,00 ~7.1149o1

Table 7. Cont. Catalog No. Description Price Each 33-66-20 B&L Microdosimeter Rods, package Of 100, silver-activated phosphate glass, 6mm long, lmm diameter, ground and polished both ends, (rods furnished with nomial diameter of.99 or 1.01 mm +.01lmm)... 44.oo00* 33-66-00 B&L Microdosimeter Low "Z" Rods, package of 100, special silver-activated phosphate glass 6mm long, lmm diameter, ground and polished both ends, (rods furnished with nomial diameter of.99 or 10Olmm +.01mm)... 44.00 33-29-40-02 Spectronic 20 Colorimeter, 340-650mu for 60 cycle 115 volt A.C., 1/2" test tube adapter, directions for use and plastic dust cover... 255.00 33-29-10 Constant voltage transformer (for unusual voltage fluctuations)... 40.00 33-29-23 Holder, for cobalt dosimeter glass, fits standard 1/2" test tube adapter... 15.00 33-99 Filter, 0.9 sptical density... 2.50 33-33-68 Extra 6-volt lamp... 2.20 *25% quantity discount on 10 packages or more. 33-1/3% quantity discount on 100 packages or more. 45% quantity discount on 200 packages or more. Procedure After irradiating the 15mm x 6mm x 1.5mm cobalt glass in known fields for doses between 104 and 5 x 106 rads, the change in optical density of the glass is measured on the Beckman DU spectrophotometer at a wave length of 340 mu and a slit width of.4mm. After waiting for a week, measure the optical densities of the samples a second time and follow this with another measurement after an hour of anneafing at 150 degrees C. An attempt at erasing the color centers should then be made at 500 degrees C. Questions 1. From the log-log plot of dose versus optical density what can be said about its functional relationship? 7.11.2

2. From the results on the annecling and the attempt at erasing the color centers, give an explanation of the mechanism of arnealing and erasing. References 1. Chapter 3 of "Radioisotope Technology"by Lo E. Brownell, University if Michiganr-. 2. Kreidl& Blair, "A System of Meya roentgen Glass Dosimetry," Nucleonics 14, No. 1, January 1956 3. Davidson, Goldblitha & Proctor, -"Glass Dosimetry," Nucleonics 14, No. 1, January 1956 4o Kreidl and Blair, "Recent Developments in Glass Dosimetry," Nucleonics 14, No o 3, March 1956 '? ~l o 3 ~

Chapter 8 Nuclear Radiation Detection and Measurement By Geza L. Gyorey and Philip R. Pluta 8.1 Discussion This chapter is designed to acquaint the student with the basic instruments and methods used in nuclear engineering for the detection and measurement of nuclear radiation, and also, to acquaint the student with the basic standards for protection against radiation. The experiments which are performed involve the use of radioactive materials of a variety of types and activities. The use of radioactive materials involves some dangers which are very slight as long as certain precautions are observed. If precautions are not observed, the danger increases by orders of magnitude. For this reason, the rules which are set forth below are enforced most rigorously in the Nuclear Measurements Laboratory. Disregard of these rules may result in the loss of privilege to use this laboratory. The reason for the establishment and strict enforcement of these rules is partly a concern for the safety of the individual student and a desire to instill good working habits. To a great extent, the reason is a concern for the safety of the public, including other persons working in the vicinity of the laboratory. The following reference material is required reading for students in this course: Code of Federal Regulations, Title 10 - Atomic Energy, Chapter I - Atomic Energy Commission, Part 20 - Standards for Protection Against Radiation, including amendments published through January, 1961 Film badges (or pocket chambers) must be worn in this laboratory at all times. The film badges may not be worn in other laboratories without the direct permission of the instructor. 8o2 General Laboratory Procedures In all work always avoid exposing yourself and others to a higher radiation dose than absolutely necessary. Be aware of the radiation field 801

intensity in which you are working. Always have a survey meter ready when removing a high level. source from.its shiel.d, and do not leave a high level. source unshielded unnecessarily. Do not handle sources of higher tthan a few microcuries of activity without the direct permission of the instructer. If you suspect that a source might have broken open resulting in radioactive contamination, speak to the instructor at onc'e. If you are not sure of what you are doing, stop and think. Do not smoke, eat or drink in the laboratory. Radioactive materials may be ingested thibs way. Be very careful when working with high voltage supplies, the shock they deliver may be dangerous. Keep the laborat;ory neat. and orderly. This promotes safety to a great extent, Cardinal. Rule: 'h.eck your hands and shoes for possible radioactive contamination when. leaving the Laboratory I:f you detect any activity, do not leave, call. the instructor. The material. in this chapter covers the subject of nuclear radiation detection and measurement, in an introduct.ory fashion. In.tensive periods of lecture, group discussion and experim.entation including a thoughtful writeup of each new technique learned are necessary to transform a novice into a confident. pract-ittione-r. of the art.. The material presented here would serve as part of the helpful back-ggroun.d material which might be read. The questiorns at t;he end o.f the chapter are typical of some of the problems which should be understood befoire- the eqluipment can be used with insight. 8.3 List of Recommended Equipment Frequently, questions are a ked. about, t- he prices of the pieces of equipment used, and also abou.t the to;al cost of the equipment in. a nuclear measurements laboratory we.ll. equipped for teaching purposes. The subsequent list is intended to answer some of these questions. It is felt that this amournt of equ.pm.en.te is quite adequate for a laboratory designed to accomodate at one time, eight to twel.ve students work.ing in fou.r groups. The equipment was ',:l.ct..led with ithose experiments in mind which are outl.i.ned in this chap'ter. From a study of the experiments 8~2

one might observe that it is quite possible to get along with considerably less equipment than the list shows. This observation is quite correct as long as all the equipment works properly at the same time, or if it is possible to repair the equipment between laboratory sessions, which is not generally true. The prices listed were taken from 1960-61 catalogs, and they pertain to high quality equipment which, in the opinion of this instructor, will excellently fulfill the requirements of a graduate level laboratory. For top quality research equipment one should expect to pay more, and conversely, lower quality equipment is available at lower prices than those indicated.. Unit Total Quantity Description Cost Cost 1 Laboratory monitor, giving audible signal, equipped with very thin window Geiger tube. 370 370 1 Geiger type portable survey meter 300 300 1 Juno survey meter 325 325 1 Radector survey meter 300 300 1 Cutie Pie survey meter 300 300 1 Slow-fast neutron survey meter 700 700 20 Self reading, neutron and gamma sensitive pocket dosimeters 45 900 1 Charger for pocket dosimeters 50 50 2 Electroscopes 125 250 1 Large air equivalent ionization chamber 100 100 1 Electrometer for ionization chamber 400 400 3 End window Geiger tubes 60 180 3 Geiger tube mounts and shields 200 600 8.3

Unit Total Quantity Description Cost Cost 2 Windowles s flow counters 200 1000 2 Gas cylinders with pressure regulators 100 200 2 BF counter tubes 125 250 2 Scintillation probes with preamplifiers 400 800 2 Scintillation probes with cathode followers 400 800 2 NaI well type crystals 1 3/4" x 2" high 300 600 1 NaI solid crystal 1 3/4" x 2"-high 250 250 2 Beta scintillation crystals 120 240 2 Alpha scintillation crystals 70 140 2 Scintillation counter shields 750 1500 7 Scalers with preset timers and 2500 volt high voltage supplies 950 6650 2 Cathode follower preamplifiers 110 220 2 - Linear amplifiers with integral discriminators 500 1000 2 1500 v super stable high voltage supplies 470 940 2 Non-overloading amplifiers with single channel pulse height analyzers 900 1800 2 Count rate meters 525 1050 2 Pulse generators 300 600oo 2 Cathode ray oscilloscopes 1250 2500 2 5 to 50 millicurie level gamma sources 150 300 1 Lead shield for gamma sources 100 100 8.4

Unit Total Quantity Descriptions Cost Cost 1 1 curie Pu-Be neutron source with shield 525 525 18 Small alpha, beta, and gamma sources 20 360 Assorted foils 100 50 Lead bricks 6 300 Miscellaneous equipment, such as spare tubes, tools, voltmeters, absorbers, source preparation kits, cables, connectors, instrument racks and cabinets, etc. 3000 TOTAL..... $30,000 8.4 Introduction to Theory of Instrumentation "Nuclear Radiation Detection" by Price is the basic text used in this laboratory course and will be covered, in more or less detail, in its entirety. Other texts which will be referred to often are "The Atomic Nucleus" by Evans and "Vacuum Tube and Semiconductor Electronics" by Millman. We are concerned with the detection of particles of interest to nuclear engineering research and development. These particles are emitted from a nuclear reactor and from naturally and artificially induced radioactive substances. Every fissioning reactor liberates neutrons, beta particles, gasa/rays and alpha particles. Raddioactive substances emit neutrons, betas, gammas and alphas, also. Of lesser importance to our studies are other particles such as protons, deuterons, positive electrons or positrons, neutrinos and the host of particles of a more transient nature such as the mesons, etc. The basis of detecting these particles is through their interaction with matter and the measurement of the disturbance caused by their interaction in the form of a current flow, voltage drop, voltage pulse, darkening of a photographic plate, change in color of a substance or its physical properties, etc.

I. tshe laboratory, we shall be most concerned with electrical methods. Charged particles are the easiest to detect electronical.ly. since they create ionization, ioe. ion pairs. in the media through which they pass -through' interaction. o.o the coulomb fiel.ds. As the particle produces ionization., its energy decreases unt;il it- no longer can supply the required ionizat-ion energy. For. an alpha particle, classical theory of radiation interaction. can quite accurately predict, f+15.s the rate of loss of energy.as a function of distance travelled, x. in the media. The result of such analysis (a derivation. is given at; the end of this section) is, dE 4 z2e 4ZNB *dx mv2 where E is the instantanneu.s energy of the. alpha associated with its kinetic en.ergy, 1/2 mv2, z e is the charge on the alpha (z 2), Z is the atomic number of the media atom, N is the concentration of media atoms and B, depending on v and Z in a logarithmic man.ner, is (see Evans for relativistic formulation for 'B) known as the "stopping number"T, From the equation. the rate of energy loss depends directly on the number of media electrons per atom, the atomic.oncentration and inversely on a energy. The number of ion pairs produced per unit path, length, of the a is known as specific ionization. An estimate of the total number of ion pairs formed can be made if the energy required to, produce one ion pair in the media and the initial, particle energy is known. For example,. the energy required to produce an ion pair in a gas has. been measured -for air _ 35 ev and for helium 31 ev. Thus. a 3 5 Mev a produces about 1,05 ion pairs in air, Sufficient numbers of ion pairs are created by a particles to produce a measurable quantity. of charge. Because of the massive size of the a its path is relatively straight as it.oses energy in the. medi.a. Electrons are also easily deteected be cause of their charge of -eo Electrons lose energy!.primarily by inelastic collision and secondarily by radiative _loss.:The ratCeof energy.loss due to inelastic coll.ision is given qualitatively by the equation for a. energy 1.oss however -the stopping number is approximately constan:t for electron energies less than. 1 Mev.- The radiative loss occurs when -the electron is in the Coulomb field Qf an atom and. is accelerated due t.o. its small mass. From classical theory9 a -charged particle emit;s. electromagnetici radiation when subjected to an acceleration. For.3 energies.less than 5 Mev the radiation effect is not of major importan.ceo I ot ice that- since - dE is pro:portional to dx z2/E the electron will travel a glreater distance in the media than an 6 8Q6

COMPTON RECOIL PROCESS.Y RAY OF LOWER ENERGY PROCEEDS IN NEW DIRECTION * ELECTRON IS EJECTED WITN THE ENERGY DIFFERENCE PHOTOELECTRIC PROCESS a. RAY COMPLETELY ABSORBED. ELECTRON EJECTED WITnH RAYA ENERGY MINUS BINDING ENERGY * PAIR PRODUCTION PROCESS. Y RAY ANNIHILATED. ELECTRON AND POSITRON CREATED AND SHARE YRAYV'S ENERGY MINUS 1.02 Mev Figure 8.1. Principal Gamma Ray Interactions.

before dissipating its energy. Gamma rays interact with matter in three important ways, photoelectric Compton scatter, and pair production, See Figure 8.1. In the photoelectric effect, the y loses its entire energy to the atom which then transfers this energy to an inner electron which is ejected. The kinetic energy of the ejected electron is equal to the energy of the y minus the binding energy of the orbital electron. Compton scatter describes the interaction between a y and an electron. The y transfers only part of its energy to the electron. The amount of energy transferred depends on the angle between the direction of the struck electron and the incident y path, and the incident y energy. For a 180~ recoil interaction, maximum energy is transferred. Klein and Nishina have calculated the differential. scattering cross section for this interaction quantum mechanically and found it to be proportional to NZ NZ. In pair production a high energy y, E > 2 moc2 > 1.02 Mev, is hv E converted in the presence of a nucleus to a beta and a positron. For intermediate energies 0.5 < E < 5 Mev the Compton effect is predominant. Since neutrons are uncharged, measurement is usually made indirectlyo The neutrons interact with matter to produce detectable radiation. Two such systems easily envisioned are a thermal fission detector and a thermal boron detector. In the fission chamber the neutrons cause fission and the resulting charged particles are detected by their ionization. In a boron detector, one has the (n., a) reaction B10 + n + Li7 where the ionization due to the a and Li is measured. 8.5 Derivation of Charged Particle Energy Loss in Matter A derivation of the energy loss of a charged, massive particle moving through matter, is included to give insight to a widely quoted result. Consider Figure 8.2, A Figure 8.2. Geometry for energy loss derivation

For the electric field in an xo cylinder of radius b, with charge Z'e inside, on the axis, recall Maxwell's equation: V D = P e (rationalized mks) rb2z V D dV P dv = Z'e I volume volume by Green's theorem, convert to a surface integral, (V ~D) dv FD n ds volume surface Z'e = Eo En (2 b) dz 3 surface Where D = ~ E and En is the normal component of the field En (z) dz = Z'e E 2itb ' 00 E (t) dt E dz be wr00 -00 where V is the speed of the charged particle. The normal impulse given to an electron in db about b is simply co00 /o00 Fn (t) dt e En (t) dt= impulse *00 Q-00 The impulse is defined as the change. in momentum of the electron,.'. assuming the electron to be initially at rest, the total energy transfer to the electron is 8~9

p2 (Impulse) 1 Z 'e2 2 mo 2 mo 2 m 1 ~,2-bT where m, is the electron rest mass. Consider a collection of uniformly distributed electrons with concentration = NZ electrons/m3) N being the atomic conco and Z the atomic no. of the media. The differential no. of electrons per unit length in a shell bounded by b + bd and b is, +(b + db)2 - i b Z = 2?Tb NZ db The total energy loss by the charged particle to the electrons in a volume between bmin and b ma is simply, b max. 2 -Z'e2 2jtb NZ db 2 4 dx?-mi n ( bm- In putting in the form of the previously quoted result, let Z'I Z, and since Price's result is in. Gaussian or cgos.o units note the extra factor 4e02 due to the mks9 and let b in max....+ B bmin 8.6 Ionization Chambers Alpha and beta particles and gamma rays interact with matter to produce free electrons and positively charged ions. In the presence of an. electric field. the electrons are forced toward the anode and the positive ions to the cathode. When the interaction takes place in the gas phase, the majority of recombination events between electrons and ions producing a neutral atom take place in the immediate vicinity of the electrode, surfaces where the concentration of atoms is greater than in the gas. In general) a measuring system must be set up which detects the effect of the ions collected at the electrodes, either in the form of an electric current voltage pulse or reduction in charge. 8.10

8.7 Mean Level Chamber A mean level chamber measures the electric current directly, see Figure 8.3, or the integrated effect of a current by means of a change in the amount of charge stored on a capacitor. In an ionization chamber, the maximum charge collected is equal to that created by the incident radiation, the primary ionization, in the absence of recombination. Gas multiplication, i.e., the ratio of collected ionization to primary ionization, is less than or equal to unity, as opposed to multiplications of 102 to 1010 for proportional and Geiger-Muller chambers, respectively (discussed more fully later). As a result, the total charge collected for each ionizing particle is very small. Leakage currents in the insulator between the electrodes must be minimized. Leakage occurs primarily on the surface of the insulator and is reduced by the manufacturer in a variety of ways. Guard rings are used in the chamber to help eliminate field distortion due to ion collection on the insulator, to define the active collection volume and to reduce leakage currents. Guard rings are kept at approximately the same potential as the collecting electrode (the output of the collecting electrode goes to the measuring circuit). A typical mean level ionization chamber useful as a laboratory survey meter is shown in Figure 8.4. Instead of measuring the mean current directly, the chamber can be used as a current integrating capacitor A = 1 i dt, i.e., the change in the voltage between electrodes can be used t8 infer the charge collected. This type of chamber is used primarily as a personnel radiation monitor. The Lauritsen chamber (Figure 8.5) has a built-in microscope to view the amount of deflection of a quartz fiber which decreases as the charge on the electrode decreases. A pocket dosimeter is shown in Figure 8.6. It is inconvenient to have a chamber large enough so that all the "associated corpuscular emission" resulting from the incident radiation is collected. This complication can be neatly avoided by using either an air wall or an air-equivalent wall chamber. These chambers compensate for the ion pairs that are produced outside the collecting region by having the ''walls" of the collecting volume made of air, or a substance whose atomic number is equivalent to air, about 8. In this fashion, ideally, the same amount of ion pairs enter the sensitive volume from the walls as are produced outside this volume. The walls of a standard air wall chamber are defined by guard rings. Bakelite and other plastics are usually used as construction material 8 ll

8.12 - IIIL~ll CURRENTI =ie ~~ II+ AMPLIFIER/ CHARGED PARTICLES ION PAIRS NORMAL ATOMS.INCOMING PARTICLES IONIZE ATOMS * ELECTRODES ATTRACT IONS. ARRIVAL OF IONS CONSTITUTES CURRENT CURRENT IS MEASURE OF PARTICLES Figure 8.3. Ionization Current.

8.13 Holder for "Set zero' filament cell control Ionization Electrometer oat val ve (E l959 chamber Window to Insulator /admit /-rays 0-25.jA meter rubber mount ings — Y.. x j ~~~~~~~~cover Control operating Tags for Grid Collecting range changing electrical resistor electrode contact components /9-volt anode battery 45-volt chamber polalrizing batt e ry Scale 1 0 1 2 3in Figure 8.4. Typical Ionization Survey Meter. (Courtesy of Atomic Energy Research Establishment)

8-14 CHARGING IONIZING BUTTON METAL RADIATION BUTTON I WIRE TRANSPARENT WINDOW FOR LIGHT METAL CAN f ~ _._ CROSS - FIBER PLATED QUARTZ FOR SIHTIN FILER * CHARGE BENDS FIBER FROM WIRE: IONS PRODUCED IN GAS BY RADIATION. IONS MOVE TO WIRE AND REDUCE CHARGE. AMOUNT OR RATE OF IONIZATION OBSERVED WITH TELESCOPE Figure 8.5. The Lauritsen Electroscope. Quartz Fiber Electroscope Eypiece Reticl Objct~e Coll/cting Eleclrode Outer Electrod e Insulator Chorgin \ D/ophragm Cap. Figure 8.6. Pocket Dosimeter.

in the walls of an air-equivalent chamber The inner surface of the plastic is coated with carbon to make it an electrode. Usually a standard known source such as radium is used to calibrate an ionization chamber. The radium is packaged to ensure that only y may impinge on the chamber. The y energy spectrum from radium and its decay products are well known, and by an extension of the arguments given in a previous chapter on dosimetry, the theoretical dose in roentgens produced in the chamber can be calculated. 8.8 Pulse Ionization Chambers The measurement of individual voltage pulses requires an understanding of the dependence of the electrode voltage on the system parameters and the manner in which the voltage pulse reaches the electrodes by induced charge build up. It is useful to define two quantities in an attempt to formulate a theoretical equation describing the motion of the charged particles in the sensitive volume. The mobility, p., is IVP where v is the average charged particle. drift velocity due to the electric field and P is the gas pressure. The mobility is different for each different kind of charged particle, depending on the chamber gas and the mean free path of an ion in the gas. Massive positive ions have much smaller mobilities than electrons. As the gas pressure is increased, the mean free paths decrease giving a reduced drift velocity as implied in the definition. The diffusion coefficient, D, is defined in a manner similar to reactor engineering, e.g., for negative particles, J-/V n_ where J- is the current per unit area due to a gradient in the negative particles whose concentration is n To gain an appreciation for the behavior of the charged particles in the chamber, write a balance equation for the positive and negative particles. Write the balance equation for electrons,. realizing that the positive ion equation is similar with the exception of signs. Consider a differential element of volume in the gas, Adx. The net rate of change of electrons is determined by the rate of recombination, leakage current due to a gradient, the current due to the field, and the rate electrons are created by either primary ionization or gas multiplication. Consider one dimension for simplicity. The balance on negative charge in the x direction is (See next page for FPigure 8.7) e Adx = Ix A + S Adx- R Adx - I+dxA where S is the volumetric rate of creation in dV and R is the volumetric rate of recombination. 8.15

Figure 8.7. Diagram for Charged Particles in Ionization Chamber. anode cathode I Idx I~ ~Ix+dx x x=6 - dx H an I D e I+v n e x ax x (the minus sign because electrons flow down the concentration gradient) Ix+dX Ix + a (I ) dx xsdx x ax X or, 2 ora _ Re a-D_ an S n atL ax ax - P ax x -ax & X where vx Usually it is assumed that D = 0, giving in genera d P acoordinatesx x. eralized coordinates, e S - R + e D n (n) The form of the recombination term is usually arrived at empirically2 e.g. away from the electrode surfaces R is often represented b.y, Constant x (conc. of electrons) x (conc. of positive ions). The balance equation has not been solved analytically. in a manageable form, for the general case. Price gives approximate expressions (pg 72, 73) which estimate the fractional loss in the saturation current (all charge collected) due to recombination and due to diffusion for certain conditions at mean level, but nothing for pulse operation. There are two main conditions required for an ionization chamber to count individual pulses, rather than a mean level current. The voltage pulse height must be separable from a low level background, and the system 8 16

must be fast enough to differentiate between individual successive pulses. To estimate the energy absorbed in the chamber gas, it is necessary to have a known relationship between the size of the pulse and the number of ion pairs formed. The time required for the pulse to build up to peak voltage and decay is important, also. Because the positive ions are generally more massive than the negative ions (usually electrons), it takes them longer to migrate to the cathode (on the order of a millisecond) whereas the electron transit is usually measured in microseconds. The following simple example may illustrate a few points. Consider a parallel plate chamber operating at 200 volts with a plate separation of 2 cm. Let a positive ion having a mass of 8 amu. and an electron be created in the center of the collecting volume. Recall,. = vP/A In air, the mobilities are approximately 16 ( cm ) (rmmOf / volt pt -_ 1070sec. cm = 1070 " giving, v_ = 1.3 x 10 cm/sec = 1.4 x 102 cm/sec Thus, the transit times are tf = 7.7 x 10-6 seconds and t = 7.2 x 10-3.- f+ seconds. These collection times on the order of a microsecond and millisecond are quite representative for most ionization chambers. A discussion is given in the next section concerning the calculation of induced charge and voltage pulse forms as seen by the amplifier. On the basis of what has been shown above, one can conclude that any ionization pulse chamber whose operation depends on measuring the effects of the positive and negative ions, the so called ion pulse chamber, must be restricted to low count rates, on the order of 100 per second because of the long time required for the positive ion to reach the cathode. This limitation can be overcome by making the RC time constant of the system longer than tf_ and shorter than t+, making the voltage pulse output less dependent on the induced charge of the positive ions. This type of chamber is the commonly known variety and is called an electron pulse chamber. 8o!7

809 Input Voltage Pul.se Forms to a Counting Circuit Caused by the Production of Ion Pairs Between Two Electrodes The electrostatic potential is a scalar point function. The potential at a point, P, which is at distances ri from charges qi, due to qi is n V(P) - K qi/ri i l where K is the dielectric constant and n is the total number of charges. The mutual energy of any system of charges can be found from the definition of the electrostatic potential. For a charge qcj. n Wj q v = i qi/rij ~ ij i=l where Vj is *the potential at the position qj due to the charges qi. The work to put all the charges in place is s.imply n n i~l j=l the factor 1/2 takes account of the fact. that the jth charge is brought into the field of the ith charge and the ith charge is brought into the field of the jth charge in the summation. Since V. is the potential where qj is located, this may be written n j l W. V... If all the charge is on one conductor, at, a potential Vo and the sum of the charge is Q, conductor 2 c qj Vj - Q2 Vo cond. But C = Q/Vo, therefore, W o =d 1/2 CV 2, which is the energ.y of a charged conductor. For a capacitor whose plates are at V' and V2 with 1 2 charge Q and -Q, respectivelyy 8 -1.8

and for a condenser, Q = C(V - V2) W C V - V 2 Il-2 Consider charges q, q,... q at positions where the existing potentials due to the other charges are V1, V2, Vn, and a new set of charges, at the same positions, ci', q2.. n' causing analogous potentials V1 ' V ' V ' By definition, 2' n n Vj =1 qj/ri. i #j J K, i= and n J K j 'ij i — =1 Multiply Vj by q.' and Vj' by q and sum over j, J n n n n n. n Vj = qj qji.qt/i v ' qj = q qi '/ j =1 i=l j=l j=l i=l j=l Since j and i are just summation indices, it is clear that n n VI j' = j' j=1 j=l This is the Green reciprocation theorem. This theorem can be extended to n conductors of potentials Vj carrying charges qj by combining the points of equal potential into a single term. Now consider a charge, q, in the vicinity of a conductor connected to a sink or source of electrons, such as ground or a battery. The charged particle produces a potential at the position of the conductor. A positive charge raises the effective potential of the conductor. Current flows from higher to lower potential, i.e., electrons flow from lower to higher potential. Thus, the effect of the charge on the conductor is to raise its potential and cause electrons to flow to it. These additional electrons may be classified as an induced charge. Thus, a positive particle induces a negative charge on a conductor and a negative particle induces a positive charge. If a charged particle is placed in the middle of two conductors, the net effect on the potential drop between the conductors is zero, since both potentials are changed in equal amounts. 8.19

However, as the particle gets closer to a particular conductor, the net effect is no longer zero, since the potential change is not identical, as is shown below. As a special case, consider two grounded conductors with a charge q at some point, P, between them, An induced charge will appear on the conductors. We can find the induced charge Q on either of the conductors, if the potential, Vp' is known, to which the point P would be raised9 q being absent, by raising the conductor to a potential Vv, using the reciprocity theorem. Recall., n n V. J - j lj j1l Giving, V q + V q = V q + V' q 11 22 1222 For the sake of clarity, let q= q and Q = q, then V = V, and V = 0 (since the conductors are grounded, giving a zero potential when Q 0). Similarly, let 0 = q and Q' = with Vp 1 and V tn w have V ' q + V'Q V p(0) + O(Q') 0 or, Q = -Vp q/V' Realize Q can be induced on. either conductor. The total charge induced on both conductors is equal to the negative of q when one conductor encloses the other, -q = Qa +Qb In the absence of a charge, q, between the conductors, a and b, at conductor volt ages of VaI and V let the potential at P be V Using the reciprocity theorem, for grounded conductors (V1 V2 - O), gives QaVa' + Qbb + q Vp 0 Substituting for Qb in terms of Qa and q gives Qa - (Vb' -V J')-/(Va - Vb'V) and =Qb -- (Vaa -: Vp)q/(Vb' - Va')

Vp' is just the potential distribution in the space charge free region between two electrodes. For parallel plate electrodes, 6 OP Va b V >V x _b a d Figure 8.8. Charged Particle Between Two Electrodes The potential between the conductors is given by the Poisson equation d' = 0; V' = A x + B dx2 when x =, V' = V x = d, V' Vb therefore, V'(x) _ b - x + Va' = Vp (x) d a Substituting, Qa g and Qb =- a = a = For a cylindrical electrode arrangement, the field is in the form, & (\) = (Vb' - Va') /r Bnr(_) where r2 and r1 are the inner radius of the chamber and the outer radius of the anode, respectively, gives, for Vb' > Va' do' = V t V, Vb ' - VaT r r.rI VbVb - V Vb ' n (a dr/r 8 r p: ro 8 21.....Vb in b~~~~~~~~~~~~~~~~~~~.. r

jr2N Q b - q (r2' q/,n zr/r) Q - Q - an q/n (rr In practice, both electrodes are not grounded The argument used to justify the application of these results to finding pulse forms is plausible, ie.,, the voltage change on the grounded or collecting electrode is small since the capacitance of the system (sum of chamber and input circuit capacitance), C, is large and for the other electrode (held at a minus potential in steady state) the effect of the induced charge is superimposed on the steady state charge and therefore may be treated separately. As a function of time, the voltage to the input circuit due to the net charge induced by positive and negative particles on the COLLECTING electrode, Q(t), since current flows in the opposite direction of electrons, the voltage of the collecting electrode drops, is, v(t) - Q(t)/C Let - Q+ (t) be the induced charge due to positive particles and -Q,(t) due to negative particles. The net induced charge is Q(t) -Q+(t) - Q (. ) V(t) =Q+(t) I Q,(,),]/J When the negative particles reach the collecting electrode, Q(t) +Nq = -Ne where N is the number of pairs created per incident radiation particle. When the positive particles are collected on the other electrode, Q+ (t) = 0, due to cancel.la-tion. with the induced negative charge. Thus, from t= - t to t+ v7(t) '= [Q.(*t) - Ne]/C 1 j~~~~o ~~e 09 V(t > t+ The voltage pulse to the input circuit; is determined by the in — duced charge on the collecting electrode.. Forthe parallel plate chamber shown above, where the collecting e.lectrode is at VBb, (t ) L= r dNe 7 J — 8 o22

For the cylindrical chamber, with the collecting electrode at the center, using Qb' V(t) = 1 n ( n Ne)e/n( C r+ N/ - The position, x or r, as a function of time is found by solving the balance equation on positive and negative charge given in a previous section. For an electron chamber, i.e., t_ < RC < t+ and x+ = xo - v+ t, x_ = xo + v_ t or r+ = r ++ v t, r_ - ro - v_ t where the subscript zero refers to the creation location and v is the speed of the particle as a function of time, we can neglect the effect of the velocity of the positive ion on the pulse from (v+ t << x or ro) giving V(t) = - Ne v_ t/Cd and, v(t) = -Ne ~n (ro/r-) C ~n (r2/rl) 8.10 Experiment On the Use of Ionization Chambers Procedure Before attempting to use an instrument, please read the appropriate sections of the instruction manual of that instrument. 1. Check the calibration of all the survey meters in the laboratory with the radium source of known strength. If you find that some of them should be recalibrated, please call this to the attention of the instructor. 2. Using the same source, calibrate one of the Lauritsen and one of the Landsverk electroscopes. Determine whether their scale readings are linear. Determine also whether the instruments are directional. 3. Determine a safe working distance from the cobalt-60 source. Determine the effective source strength of this source in millicuries. Calculate 8.23

the effective strength of this source at the time of its purchase. 4. Charge a few of the Victoreen and a few of the self-reading pocket chambers. Expose them in a known radiation field for a known length of time so. that they accumulate a radiation dosage of about 50 mr. Compare their readings with the calculated results. 5. Compute the expected relationship between the radiation dose rate and the meter reading for the G.E. air equivalent ionization chamber used with an appropriate electrometer. Decide on which of the two sources mentioned previously you want to work with. Make measurements with the chamber for at least two different source-to-chamber distances. Use a collecting voltage of 135 volts. Compare your data with the calculated results. At the smallest of the source-to-chamber distances take readings also for collecting voltages of 90 and 180 volts. Note and explain any variation in the results. At the largest of the distances connect one of the guard rings directly to the collecting electrode, and compare the resulting reading with that; obtained when the guard rings were connected properly. Explain the difference using a schematic circuit diagram of the apparatus. Determine how much of this difference is due to ionization near the end of the chamber and how much of it to leakage through the insulation. Measure roughly the time constant of the apparatus o Please include in your report your calculation of the relationship between source strength and dose rate for radium and cobalt 60. Explain what is meant by an air-equivalent ionization chamber. 8.11 Proportional. Chambers Proportional counters are similar in principle to the ionization chamber and the Geiger-JMuller counter, in that all three operate on the principle of gaseous ionization. In. general, the proportional counter operates at a higher voltage than an ionization chamber. By using a fine wire as the collecting electrode, it is shown below that electrons gain sufficient kinetic energy from the field to cause secondary ionization, et, in the immediate vicinity of the anode, see Figure 8~9. Over a limited voltage range, the proportional voltage region. gas multiplication is independent of the amount of primary ionization. With large gas multiplication, a preamplifier is not required. The multiplication depends primarily on the magnitude of the electric field. Since this dependence is very sensitive to small fluctuations in the collector voltage, the proportional counter is of little value as a mean level chamber, and is used almost exclusively as a pulse chamber. As a pulse 8 24

chamber there is a discrimination between incident particles which produce different amounts of primary ionization, because of the constant gas multiplication. For example, a proportional chamber can be used to differentiate between incident a, % and y, see Figure 8.10. If most of the gas multiplication takes place in the immediate vicinity of the anode wire, the total charge produced is almost independent of where the primary ionization originated in the chamber. The above may be verified as follows: To compute the field, assume that the charge density is zero between the electrodes, then V * VV = 0 Using the boundary conditions that V = Vo at r = r (cylindrical coordinates, r = 0 at the center of the anode) and the inner surface of the cathode is at r2, recall that V= - aV/6r. Assume angular symmetry r V + d V r d= + 0 2 dr2 dr dr dS = _ dr r s = K/r i O;r2 V, dV -K! dr K = V0 Jrl rn r2) / rlI giving, = Vo/[r gn( )] Quantitatively, one asks at what space position in the chamber will an electron have gained enough energy to create an ion pair. In air, this takes about 35 ev. An electron has an interaction on the average every mean free path. Since the interaction is primarily with the orbital electrons of the neutral gas molecules, it is reasonable to assume that the energy transfer in the form of thermal excitation is large enough to prevent the creation of a secondary electron, unless enough energy to create a pair is gained during a single mean free path. The mean free path. The mean free path for electrons in air is about 2 x 10-4 cm. at 1 atm. Let rl = 0.001 inches, r2 = 1 inch, Vo = 500 volts. The field required to 8a25

8.26 PRIMARY, IONS - 0 PULSE Q, o ~ - Ljili COUNTER ~ A /. /. ELECTRONS RECEIVE ENOUGH ENERGY TO IONIZE - AVALANCHE OF SECONDARIES CURRENT MULTIPLIED BY 1.000 TO 1.000.000 Figure 8.9. Simplified Representation of Gas Multiplication. ":;''PULSE i:;1'G-M REGION: RECORDER,. PRI2M METHANE (I0' PRIMARY PAIRS) l; I_ * 2 O 10 8 VOLTS "600 GCAS-FLOW TYPE. PULSES PROPORTIONAL- TO IONIZING POWER. RECORDER CAN BE SET FOR PULSE SIZE RANGE,CAN COUNT o'S IN LARGE FLUX OF- 3'S. NO WINDOW TO STOP PARTICLES Figure 2.10. A Gas Flow Proportional Counter.

3/5. 2\ 4 create a pair is just 3= 17.6 x 10 volts/cm. This field occurs in the chamber 2 x 10 at r = 500/(17.6 104)>n 103 or, r = 4.1 x 104 cm but r1l is 2.5 x 10-3 cm, so must either up the voltage or lower the gas pressure, thereby increasing the mean free path, to get multiplication. Suppose the voltage were raised to a very high 2 x 103 volts, find r = 1.6 x 10-3 cm, which still won't give multiplication. Hopefully, the above has illustrated why proportional ( and Geiger Muller ) counters are almost always operated at reduced pressures. If instead the chamber had been at 500 volts and 10 cm of Hg, (? = 3552/[2 x 10-4 76 )] = 2.3 x 104 volts/cm, and r = 3.1 x 10-3 cm. It has been found by experiment that the gas multiplication is less sensitive to small voltage changes if a polyatomic gas is added in small amounts to the chamber gas. It is thought that the polyatomic gas reduces photoemission by absorbing the electromagnetic radiation. For pulse work, the details of the pulse time behavior are important. Since most of the ion pairs are created next to the anode, there are a large number of localized positive ions surrounding the anode after each avalanche. The electrons are collected quickly because of their higher mobility and proximity to the anode. The positive ion sheath around the anode moves relatively slowly toward the cathode. By having the input time constant less than the positive ion collection time and greater than the electron collection time, fast pulses are produced whose heights are proportional to the primary ionization. Price plots the pulse form in Figure 5.4 as a function of the ratio of system time constant to the collection time of the positive ions. It should be noted that quite high resolving times are possible because the ion sheath is localized, i.e., another avalance can occur at a different position along the wire after a previous pulse has passed the discriminator level, and be counted, in spite of the fact that the ion sheath of the previous pulse is making a localized field distortion. For highest accuracy in measuring pulse amplitudes, pulse pile up must be avoided, limiting the count to a maximum on the order of 104 to 105 per sec. depending on chamber characteristics. Proportional counters are used widely for a particles. Because of the short range of the a, it is convenient to be able to put the sample inside the detector, right in the chamber gas. Because it is often convenient to use a differs t or a purer gas than the usual laboratory air, most proportional counters are made as gas flow counters. In the gas flow counter, the chamber gas is purged of laboratory air after a sample 8 27

is introduced by means of a special regulated gas supply which feeds the chamber. After the chamber has been purged, the flow of gas is reduced to a bubble per second or less to insure that the chamber gas will not change as a result of high level count rate operations and minimize the effects of a leak. Because of the discriminator in the counting system, proportional counters have extremely small background noise levels when used to detect a or B. The counting rate as a function of counter voltage has a characteristic plateau. For a given multiplication, particles with the largest specific ionization are first to pass the discriminator, e.g., a before P. Since a's are almost monoenergetic, and e's have an energy distribution, it is to be expected that the p plateau have the greater slope. Typical count rate as 'voltage curves are shown in Figure 8.11 for a and p emitters. Chambers in which the sample emits radiation which subtends an angle of 2 it radians in the chamber gas are known as "2 itc" counters. There are considerations such as backscatter and self absorption which require attention when the sample is placed in the chamber and one desires to measure or calculate the absolute rate of disintegration. In backscatter, radiation which is emitted originally in a direction which would exclude it from the active chamber volume suffers a number of scattering events such that its direction is changed and it does produce a measureable pulse. Self absorption simply refers to the count loss due to absorption of radiation by the sample itself. Price covers these details quite adequately. 8.12 A Note on Gas Multiplication (See Figure 8.9) Consider N electrons with sufficient energy to produce an. ion pair. Let the probability that this electron produce a pair be e. As a first approximation, assume that c is independent of the electron energy, i e., the only requirement is that the electron has the capability of producing one pair, but as it gains energy in the field it may produce one pair after another. Let us follow the creation process for n generations. Start with N, these produce NE, giving a total of N + NE; these produce (N + NE)E or a total of N + 2NE + N c2, and so on for n generations giving a -total of N + 2nN + nN2 + o + nNn-l + NEn 8 2a8

Let M1 be the total number of electrons produced per incident radiation particle. M N[l+n( 7 el +) i=l Recall, 1 1 + e + e2 +... + +ne 1 - e for e ~0DO and n -0 a,.., M = N [1 + n (-1 + 1/1-). The gas multiplication usually referred to, in the absence of photoemission, is M1/N, and is most often determined empirically. There are some additional counts introduced because of the interaction of the electromagnetic radiation associated with the production of the secondary, tertiary, etc. electrons with the neutral gas atoms and principally the walls of the electrodes. Electrons are ejected into the chamber gas due to photoelectric effect interaction of this radiation. These electrons, caused by photoemission, initiate new gas multiplication. Since it is desird'.le to maintain proportionality between the energy deposited in the gas by incident radiation and the chanber output pulse size, an organic gas is usually added to suppress this effect. 8.13 Experiment on the Use of the Proportional Counter Use the windowless flow counter in this experiment. Please avoid initiating a continuous discharge, or introducing air into the counting volume. Before starting the experiment, read the appropriate instruction manuals, and see whether the pieces of equipment are connected correctly. 1) Determine the count rate versus voltage characteristics of the counter using a thorium foil. Explain the meaning of your results. Use the oscilloscope to observe the behavior of the pulses as the voltage is varied. Determine the voltage at which some of the pulses cease to be proportional to the initial ionization. 2) Obtain a count rate versus voltage curve for background. Note the magnitudes of the background count rates in the alpha and beta plateau regions. Explain why there is a background count in spite of the fact that no alphas or betas can pass through the shield surrounding the counting volume. 8 29

3) Cover the thorium foil with some material which is just thick enough to absorb all the alphas. Obtain the count rate versus voltage curve for the covered source, Compare this with the one obtained without the cover. Comment on the similarities and differences. 4) Obtain a count rate versus voltage curve using a source emitting betas but no alphas, such as cobalt 60. Compare this with the curve obtained using the covered thorium source. Commente on the similarities and differences. Explain how a plateau in the proportional region differs from a Geiger plateau; discuss pulse height variation and the cause of the positive slope of the plateau.. 8.14 Geiger-Muller Chamber The Geiger-Muller, GM, chamber is -the simplest to understand and the easiest to use of the gaseous ionization chambers. The GM is just an extension of the proportional counter concept to higher gas multiplication, using a higher collection. voltage. The gas multiplication at these higher voltages is so great that the primary avalanche produces a cascade of daughter avalanches by photoemission which are continued until the positive ion sheath which surrounds the entire anode reduces the field below that required for gas multiplication to continue. At an. even higher voltage, the chamber is reduced to uselessness because of the essentially continuous discharge caused by photoemissiono Since the size of an output pulse on the GM plateau is on the order of a vol.t very little external equipment is necessary to count pulses. Thus, the GM tube is a low cost detectors as well as rugged see Figure 8o12. The GM is versatile in -the sense that an incident, a, B or y. etc. all produce the same size output pulse, once gas multiplication. has begun. Since only one ion pair is required *to initiate a discharge in the absence of recombination9 GM is useless. for energy resolution9 or pulse analysis The GM tube can be used for count rates almost as high as in a proportional counter. The same equations may be 'used to describe the pulse time characteristics and ion transit time with the understanding that; the massive collection of positive ions near the anode in the GM will tend to modify these reuslts due -to the greater field distortion. As in the proportional chamber, the time constant of the input circuit has a value between the collection times of electrons and positive ions for maximum resolution. 8 30

RECOMMENDED OPERATING REGION FOR COUNTING ALPHA RECOMMENDED OPERATING REGION FOR PARTICLES IN PRESENCE OF BETA AND GAMMA RAYS BETA AND GAMMA COUNTING 1.2 - - - - - - -- - -\ - - - - 1.0,.2 --- r —.-. ---- m- ~" Z=X 0 z — 600 700 800 900 1000 \1000 1200 1300 1400 1500 1600 17 00 VOLTS ALPHA COUNTING BETA S GAMMA 10 MILLIVOLT COUNTING I MILLIVOLT SENSITIVITY SENSITIVITY Figure 8.11. Proportional Counter Characteristic Curves for an Alpha, and a Beta or Gamma Emitter. (Courtesy of Tracerlab Inc.).~SIIVT ', 0oIIT <~~~Fgr 81.4 'rprtoa 'one.....rsi Cre ora lpa nda X~ ~ ~ ~~~o anaEitr.(oreyo rcra n.

8.32 G-M TUBE OUTPUT NEGATIVE PULSE H.V. Figure 8.12. Basic Circuit for Geiger Muller Chamber. UV LIGHTe POSITIVE ION DRAWS ELECTRON FROM CATHODE ION BECOMES EXCITED ATOM ATOM RADIATES IN ULTRAVIOLET. LIGHT EJECTS PHOTOELECTRON FROM CATHODE. ELECTRON INITIATES FURTHER CASCADES Figure 8.13. Need for Quencning in a GM Tube.

GM counters require quenching; see Figure 8.13o That the energy difference between a positive ion reaching the cathode and the work function of the cathode is sufficient to cause the emission of an electron or a photoelectron into the chamber gas to produce another avalanche is well established. The idea of quenching this behavior by mixing gas molecules with higher work functions in the chamber gas was the next logical step, see Figure 8,14. In the organically quenched tubes, after the positive charge is transferred to the organic molecule, the excess energy is used to dissociate the molecule when it reaches the cathode. The molecule can be used only once to transfer the charge since the dissociation is not a reversible process. This difficulty is avoided by using halogen quenched tubes in which recombination does take place. Because the ion sheath surrounds the entire anode, thereby reducing the field, for a certain fraction of the time after an avalanche has stopped, another avalanche can not be initiate because the field has not had a chance to build up to the necessary strength. The time during which the chamber is insensitive to incident radiation, as passed by the counter discriminator set for the GM plateau voltage, is known as the GM dead time. Any primary ionization during the dead time is not counted. The dead time sets the inherent lower limit at w7aiich the GM can resolve individual pulses. The dead time depends on the magnitude and space and time distribution of the ion sheath. The dead time is slightly different for every pulse counted in single tube because of statistical fluctuations in these quantities. The characteristics of the voltage -amplifier used will influence the effective pulse resolution of the system. The resolving time usually refers to the time which the system, the GM and associated pulse amplification and detection equipmen-t, requires to distinguish between two successive pulses. The resolving time is always greater than the dead time, by definition. The form of the pulse from the GM was considered in a previous section. When GM tubes are used for absolute particle counting, e.g., to determine the activity of a radioactive sample, it is necessary to correct the measured count for all losses, For a given counting geometry it is necessary to make corrections for the number of particles which don't produce a measureable pulse having reached the sensitive chamber volume, backscatter into this volume, self absorption in the media between the surface of the source and this volume including the walls of the GM tube, dead time or resolving time losses, and the number of multiple counts not produced by the types of particles being measured which originate from the source such as background, and spurious photGoemission counts, The 8e33

8.34 EXCITED POLYATOMIC MOLECULE POLYATOMIC MOLECULE DISSOCIATES INSTEAD GIVES UP ELECTRON. OF RADIATING TRAVELS AS + ION TYPICAL GAS.090% ARGON 10% ETHYL ALCOHOL OR AMYL ACETATE, ETC. NEARLY COMPLETE SUPPRESSION OF SPURIOUS COUNTS BUT... DISSOCIATION OF GAS LIMITS USEFUL LIFE Figure 8.14. Action of Quenching Gas in GM Tube.. SAMPLE 90 ~lb i /, --- o' I_____VOLTAGE NO PULSES LARGE ENOUGH FOR RECORDER AVALANCHES GROW LARGER WITH VOLTAGE NUMBER OF RECORDER PULSES INCREASES O ALL PARTICLES TRIGGER MAXIMUM AVALANCHE ALL PULSES RECORDED OPTIMUM SETTING FOR OPERATING VOLTACE HIGHER VOLTAGE GIVES SHORTER TUIBE LIFE SPURIOUS DISCHARGES LEAD TO BREAKDOWN Figure 8.15. GCM Characteristic Plateau Curves.

straightforward account by Price is self-explanatory as are the bulk of examples of tube types used for different applications. A typical characteristic curve of a GM chamber is given in Figure 8.15. 8.15 Experiment On the Use of the GoM. Chamber Procedure Please exercise due caution in working with Geiger-Muller tubes and the associated electronic equipment. The end windows of the tubes are very fragile, do not let anything come into contact with them. The high voltages used present a shock hazard. Turn off the high voltage supply while connecting the equipment. Excessive voltage can damage the tubes. Before turning on the high voltage supply, turn the voltage adjustment down as far as it will go. Increase the voltage slowly and only while the scaler is counting. 1. Determine the count rate versus voltage characteristics of an end windowtube using any convenient source. Determine the useful length of the plateau, and pick an operating voltage. Explain why the plateau has a positive slope, and justify the selection of your operating voltage. 2. Familiarize yourself with the cathode ray oscilloscope. Observe the pulses coming from the GM tube. Plot pulse height versus applied high voltage in the plateau region. At the operating voltage, estimate the dead time of the tube, the resolving (dead) time of the counting system, and the recovery time of the tube. Compute the counting rates at which the dead time (or coincidence) losses are about 1%, 5 %, and 10%. Observe and comment on the variation in resolving time with changes in applied high voltage in the plateau region. 3. Determine the dead time of your counting apparatus by the two-source method. Use the T1-204 split sources and the inactive duimmnies. Position the sources in such a way that the counting losses are between 5% and 10% when the sources are counted simultaneously. The following procedure is recommended. a) Count the sources simultaneously for 15 minutes. b) Replace source lby dummy 1 withaut touching soarce 2. Count for 10 minutes. c) Replace source 2 by dummy 2 without touching dummy 1. Count for -5 minutes. d) Replace dummy 1 by source 1. Count for 10 minutes. Calculate the dead time. Compare this with the results of the measurements with the oscilloscope. 8;35

4. Observe the effect of operating voltage on the resolving time of the system and explain. 8o16 Statistics We attempt to determine certain aspects of nuclear decay phenomena by measuring emitted radiation. The inherent laws governing the nuclear processes are statistical in nature. The distribution functions which describe the statistical fluctuations are known. The most useful distributions in our work are the Poisson, Normal or Gaussian and Interval functionso For a thorough study of statistics read Cramer's Mathematical Methods of Statistics. (Princeton University Press). 8.17 Poisson Distribution The Poisson distribution is a special case of the Binomial distribution which is applicable when the mean life of the emitting sample is long compared to the experimental observation time, and there are sufficient identical, independently acting atoms to allow significant data. In more general terms, a Poisson distribution should be expected from a random process in which the probability of an event happening is small and independent of time. The probability that an observation will result in x counts over a time interval for which the true mean count is m is given by mxe -m PX = x' (read especially Chapters 26, 27, and 28 of Evans and 3 of Price). Note that Px is not a continuous function, since x is restricted to integer values, A representative histogram of a Poisson distribution can be given, see Figure 8.16. P ___ x Fn Figure 8.16. Poisson Distribution. 8.36

The distribution is not symmetrical, but is skewed toward x < m. The assymmetry increases as x -O. The Poisson distribution describes most nuclear radiation decay events for the listed assumptions, and may be used when only a relatively small amount of data is available. For any distribution, a measure of the spread in the data about the mean is given by the standard deviation, a. a2 = Z (x-m)2 Px = the variance all x For Poisson, 2 _ z (x-m)2 mx e-m x=O x or a= Em As will be shown under the Normal distribution, about 68% of the observations should fall in the region m + a. In a practical calculation, with N finite observations, the m ca-Jn.,77 is given by, _1 N i=l and N 2 - 1 Z (xi-) sample N-1 i=l From the theory of errors, it is known in general that the distribution of mean values tends to be more nearly Normal than the distribution from which the means were calculated. Thus, it is shown that the mean found by repeating the entire experiment would fall within x + oa- about 68% of the time, where a- = ao/N. x Normal Distribution The Normal distribution is an approximation to the Binomial distribution for a large number of observations and a constant average value. The probability that x will lie between x and x + dx is given by, 1 -(x-m) 2/2a d dP = ed x 8 37

Note that a is independent of m, whereas in the Poisson, a = Tm., The normal distribution lends itself nicely to analysis and gives an exact interpretation to a. As the Poisson, the Normal is normalized to unity, 00 dP 1 This distribution is not restricted to integer values of x, and for large N, is essentially continuous,- and symmetrical. From the analytic distribution it may be shown that tangents to the curve at the points of maximum slope intersect the x axis at m+ 2a, also, the half width at half maximum is given by4f\ a. Thus, a may be found from the data in graphical form or from the variance. By direct integration of dPx it can be shown that 68% of the observations fall within m + a. dPx dx 2,Sn a m-25a m m+2a Figure 8.17. The Normal Distribution For a process governed by Poisson, the analytic Normal and Poisson distributions become identical as N —Xoo for mean values greater than approximately ten. Given a population (complete set of data) whose mean, m. and variance are known, the sample (smaller amount of data) mean, x, has( s variance. For the sample, the variance of the variance is - given by a2/2N. These results are true for all values of N for a Normal distribution, but depend on knowing the true population m and a, This distinction must be noted and any analysis of 1. or 2 data points must take account of this fact. One can check on the probability that the given sample-mean is representative of the population mean by using the student t distribution where, t.- (x - m) N / sample 8.38

and t follows the distribution, 1 (F + 1) dP = dt / aF B 1 F)(1 + )2 2, 2 F B representing a Gamma variate and F the number of degrees of freedom (see below). By integration of this distribution, significance limits can be obtained as was done, for example, with the Normal in determining the value for a, etc. The t distribution can also be used to decide whether 2 sample means differ significantly or whether they can be assumed to be from the same population. 8.18. Interval Distribution The interval distribution describes the probability that the time interval from t to t + dt between random events which have a constant mean rate, a, and a Poisson distribution is given by DP = a eat dt The number of intervals falling between tl and t2, n, is t t2 n = Na e-at dt = N[e-atl -edat2] Jt1 where N is the total number of intervals between t = 0 and o. The average interval is just t =_l/a. Letting t2e oo, the probability that an interval is longer than t is n -1 n = e = 0~37. N Thus, 63% of the intervals are less than t, an important consideration in estimating counting losses due to counter dead time. 8.19 Checking Equipment Using Count Rate Data Since the nuclear phenomena being studied follow the statistical distribution law, we expect the results of a measurement to follow the same law, neglecting counter losses. If there is no spread in the data, something is wrong. If there is considerably more spread than the law predicts, something is wrong. The initial data taken in every experiment should be analyzed to see if it falls within the range of acceptability from a statistical standpoint. The Chi-squared test is one of 8 39

the most decisive in determining whether or not the data follows a particular distribution. Consider the quantity, n N2 / (xx) 2= N - 1 i =1 For a true Poisson, N-ooaIn, Q2 = 1. For data which is not Poisson, or if an insufficient amount of data is used, the calculated Q2 f 1. It is difficult to tell the amount by which Q2 can vary from unity and still represent acceptable data. To establish an acceptable spread from the mean in a counting experiment, chi, X, was defined as follows, X2 [ (observed - expected) 2/expected]; i where the summation is over the number of discrete counts recorded in a fixed time interval. Applying this to a Poisson distribution in which a total of M time intervals were considered, the expected number of observations of a count, x, is given by, L =M P and the observed number of time intervals in which x counts were recorded is 2x, giving, 2 2Lx] X = [(2e L) LI Theoretical studies have shown that the approximation to the true chi value (oo amount of data) using a finite amount of data is acceptable if there are at least 5 discrete counts each having been recorded at least 5 times. If there are fewer than 5 discrete counts then there must be more than 5 intervals in which such counts are recorded. Define the number of degrees of freedom, F, which is the number of independent ways in which the series of discrete counts may differ from the expected series. The maximum degrees of freedom are i. For the Poisson described above, two degrees of freedom are used up, first, the total number of events (to calculate the mean used in Px), giving F = i - 2. 8040

2 The random variable, W, possesses a X distribution if the probability that W takes a value in dwi about X is given by F- 1 F PF do = [(2 )/( ( ] e do cl > O, where F is the number of degrees of freedom. PF d& = 1. For this distribution, the mean = F, and the variance o = 2F. As F increases, the distribution approaches the Normal, if the Normal standard deviation is replaced byT2F and the mean by F. For F > 30, the X2 distribution can be replaced by the Normal as a good approximation. In selecting N items from a population, the value of each of the N variables may range over all values in the population, naturally. The /2 distribution can be used to test how well a sample distribution agrees with a population distribution. Let xl, x 2... xN (N = total number of data points taken, whether or not the same7 be a sample of values of 9 and let the range of ' be divided into r class intervals Y1 < x < Y, Y2 < x < Y Y < x < Y. Let the number of values of xi from the sample falling into these intervals be fl f2. fr' respectively. Let the relative frequencies in these same intervals expected in the population distribution be g, g,... so that the number of values expected in each interval from a sample of N points are fl Ngl, f2 = Ng2 fr' = Ng, respectively. Now, r ~2 2' ~, r r f 2 2 2 X fi] i It has been shown that for a large enough N, X2. This limiting distribution is independent of the population distribution. Usually, require each fi > 10 for this approximation of "large enough N" to be acceptable. Thus, the chi-squared test compares the sample X2 -to a theoretical distribution of X2 which is independent of the type of population distribution from which the sample is taken. The theoretical X2 distribution can be integrated to obtain significance limits on 2. The results of such integrations are shown on page 776 of Evans in terms of the number of degrees of freedom and the calculated X2 with the probability, P, that larger deviations from the theoretical population distribution would be observed in the next series of data taken from the same population. The significance limits set on P of 0.98 and 0.05 are arbitrarily set to establish whether or not the data is statistically acceptable. 8.41

8.20 Degrees of Freedom The number of degrees of freedom, F, measures the extent to which fi are known in advance to agree with f v. In all cases the sample size and the number of population values useA are equal. Thus if r - 1 of the fi' are given, the remaining fi' can be calculated, therefore, F = r 1. This is the value of F to use when the population distribution 2. is prescribed and the Xs is used to test whether fi is consistent. When the population distribution is determined by fitting some distribution to the sanple, i.e., when the sample and population distributions are made consistent by means of their means, variance, moments, etc., for each such parameter, a degree of freedom is lost. For b of such parameters, F = r - 1 - b. 8.21 Test of Fit of Sample to Poisson For a sample of positive integers, xl, x2,... xN from a population known to be Poisson with an unknown parameter, X, proceed as follows: Divide the integers i into r intervals, where the first interval contains the integers 0 < i < C1, the second, C < i < C2 Ckl < i < k C the rth, Crl < i < co. Let the sample frequency of occurrence of values x. in the kt interval be vk. For Poisson, the probability that x = 1 is i i The population frequency of occurrence of values in the kth interval is k =ZNP., where the sum runs from i = C to C - 1. Thus, r = 7 (vk - k )/k k=l N The minimum X2 is assumed if = x =- x.. The approximation involved N L 0 r j=l in setting /2= 7 (Xk - x) / as done in Price should be understood. k=l 8.22 Several Simultaneous Statistical Fluctuations Consider simultaneous, independent random fluctuations, such as a nuclear disintegra-tion and background noise due to stray radiation. Let x, y, z,.e be the average number of particles from the independent sources per unit time producing a, b, c,.. specific effects per particle which are detected. The average detection rate is 8 42

u = ax + by + cz +.. For the Poisson distribution, it can be shown that the variance of a single observation of u is given by 2 2 2 2 2 = a2x + b2y + c2z + For example, if one records 100 ca particles, each producing 105 ion pairs and 10 roducin 103 ion pairs in an ionization chamber u = 2 x 107 and ca r= s1101 + 1017 = 1.005 x 106 showing that the a contributes 99.5% of the fluctuation. If a second chamber records only 100 a's per unit time and its output subtracts from the first, c = -105 and z 100, giving u = 107 and a=l.4x106, or the fluctuation is greater than for the uncompensated chamber. For a Geiger tube, where every incident particle results in one voltage pulse, a = b = 1.0. In general, when any number of random statistical quantities which are combined by addition or subtraction, the result can be given as, 2 2 1/2 (a + ~-) + (b + of) +... = (a + b +...) + ( + +..12) a b- a b =?+ C For example, to determine the difference between (source + background), (v + a_) and background, (ui+ cr) count, the best estimate is given by, v( + a)souce = (v - u) +. (a_ 2 1/2 s+ ) s = (v- u + (a2 + a_2) V U If the difference between two means is twice the larger a of the two, there can be little doubt that the means represent different quantities. This is especially important when measuring a signal which is not much larger than the background. A typical example would be an ionization chamber with a background of 5 Cl's and 50 3's per unit time, giving a =a2x + y= 2.2 x 105 ion pairs. Using the above definition for significant data, the smallest numbers of H's which can be detected is 2a/c = z, z = 2(2.2 x 105)/103 = 450 n/unit time. Realize that the mean signal to noise ratio is a useless quantity... the fluctuations in the signal and the noise must be given. It is one thing to report the results as cp = 631.6938 and quite another to- report p = 632 + 54. Never give a result without some measure of the spread in the data. In particular, I recommend that you read Chapters 26, 27 and 28 in Evans for a more detailed presentation of most of the work covered in this paper. 8 43

To multiply or divide independent results, we have, + C) x ( a_) x (.-... )+ a x b x 2 + -a - b -a and b +_ a4) b b \ ab These results may be applied again to the usual counting experiment to show the importance of proper background and source counting times in taking good data. Let the average background rate be b for a time tb and b+s be the rate with a source and background for a time ts. The background can be expressed as (btb + btb )/tb and the source and background as ([s+b]t + N (s+b)t )/ts, giving the net count rate due to the source alone as s + r b b and depending on the relative magnitude of s and b, one can choose a reasonable t and tbo If b << s then want ts > tb If b >> s then for s+ t+b require (t + tb) < 8tb etc. For a fixed counting time, ts + tb constant, Price suggests a minimum in the standard deviation is obtained by setting dor/dt =. ie,, 2 2 -2as das = [(s+b)/ts] dts + (b/tb) dtb 0, but dts/dtb = -1 giving (s+b)/b = (ts/tb) for minimum standard deviation. 8.23 Counting Loss Due to Chamber Dead Time In the interval distribution, it was shown that the fraction of time intervals associated with events which occur between time t and t is n = N (e-atl - e-at2), where a is the count rate and N the total number of intervals for all times. Consider a chamber which requires a minimum time, e, between events, whetser or not these events are recorded, before it can record the next event, The fraction of the number of intervals which are separated by times less than or equal to e is just (l-e-a). The observed number of counts is the true count (no. of intervals) minus losses due to the chamber resolving time, n = N - N(1-e-Ne/T) = Ne-NE/T 8,44

where the true count rate is N/T. The observed count rate is then, rn = RN e-RNc. This type of chamber is referred to as 'paralyzable' since the count goes to zero as the number of events goes to infinity. The other extreme chamber type can record only events which take place after a time e, after an event has occurred in the chamber. Events which occur between the count time and count time plus e do not affect the chamber in any way, and are the only events which are not recorded. The fraction of the time during which the chamber is insensitive during a total counting time T is ne/T. Thus, the fraction of the true count which is being observed must be n/N = (T-ne)/T, giving N = Tn/(T-ne). The true count rate and the observed rate are, R = rn/(l - rne) rn = RN/(1 + RNE) Thus, as N o, RN - and rn ---l/. The fact that this counter retains a finite count has resulted in the descriptive label, 'non-paralyzable '. We shall be using counters or rather chambers of both these types. The counter resolving time usually may be found from an oscilloscope trace directly, or by comparing the statistics from a counting experiment using the standard two source method at low count rates. 8.24 Experiment on the Statistics of Counting Counter Plateau Use the well counter and end window GM tube and any convenient source. Determine the count rate versus applied voltage curve in voltage steps of 25 volts. Determine the useful length of the plateau and its slope. Statistics Dead time By observation of the pulseform from the GM on the oscilloscope screen read directly the values of the dead and recovery times. Calculate the dead time from the split source data of a previous experiment. Compare with the CRO results. Calculate the statistical deviations for the measured count rates and modify the counting times used to determine the dead time for a better result. 8 L-5

Prove that the time for counting with two sources should be 2 longer than for either one source count if both sources have equal strength and background can be ignored. Calculate the standard deviation of the calculated dead time (hint, see pg 49 of Experimental Nucleonics by Bleuler and Goldsmith). Fluctuations at low counting rates Count background 30 times for an interval of 30 seconds. Let n be the counts per interval, and N be the number of trials. Determine the average number of counts per interval and the standard deviation. Derive a formula for the standard deviation of this series of measurements in terms of the counting rate and the time interval used} and the mean. Compute the variance of the data using the standard expression involving the mean, the individual counts and the number of trials. Discuss the significance of this second moment quantity as relates to our measurements. Compare the theoretical and experimental standard deviations as found by the above. Discuss the reasons for the observed difference. Determine the number of counts ni for which the absolute value of deviation from the mean exceeds 0.67, and 1, 1.6, and 2 times the standard deviation. Compare with the theoretical probabilities. Is such a comparison meaningful? Are the results you found experimentally reasonable? Fluctuations at High Counting Rates Repeat the procedure and calculations for a source giving about 5,000 cpmo Count for 10 intervals of 2 minutes each, or a more optimum time if you can determine it. Introduce dead time corrections only in the final results, not in the individual counts, Compare the percentage error of the higher and low count rates with theory and discuss. Poisson's distribution Count background for 100 intervals of 5 seconds each, Determine the number of intervals during which n=O, 1, 2,... counts were observed, Calculate the average number of counts. Compute the probabilities from the Poisson using the experimentally determined mean. Plot the experimental and theoretical probability distributions as a function of n. Discuss and explain any differences, Do a chi-squared analysis of some of your interesting data (high2 low or both types of count rate). Discuss results, 8.46

Considerations 1. Show that for radioactive decay the standard deviation can be expressed as the square root of the mean, Make very clear the assumptions and limitations placed on the physical model to arrive at such a result, 2. It is desired to male a measurement of the counting rate due to a given source in a shielded GM tube enclosure. A 1 minute count with and without the source resulted in 790 and 33 counts respectively. What would be the optimum counting schedule with and without the sample, and what would be the accuracy of the results if the total available time is one hour? 3. What is the difference between a sample and population mean? When only a limited amount of data is available, what restrictions does this place on your statistical analysis of the results? 4. Is one justified in using precisely defined quantities for a Normal distribution in the analysis of Poisson data? Discuss. 5. Does the chi-squared test have any value or worth over the more standard means of analyzing data? 8,25 Scintillation Detectors The scintillation detector utilizes two well known phenomena to count incident particles and analyze their energies. The interaction of radiation particles with certain substances results in the emission of electromagnetic radiation. This radiation has an energy which is suitable for the production of energetic electrons by the photoelectric effect. These electrons can be increased in number by means of a special electronic tube, the "photomultiplier tube", until they exist in sufficient quantity to be readily counted. If the number of primary electrons released from a "photocathode" by the electromagnetic radiation are proportional to the energy dissipated in the scintillation material by the impinging radiation, and if the percentage increase in the number of electrons in the photomultiplier tube is independent of the number of primary photoelectrons, and if the entire process of absorption and photomultiplication takes place in a short time, say less than a microsecond, it is clear that the energy distribution in the impinging radiation radiation can be deduced, The above remarks summarize the manner in which a scintillation detector works. The details of the production of the electromagnetic radiation and the multitude of different substances which can be used as scintillators, or synonomously, which can produce light energy from molecular excitation and ionization, the collection of this light in the most efficient manner, etc.,

etc., should not hide these basic concepts. Scintillators can be used to detect most particles and in fact are so sensitive they were used to measure the presence and confirm -the existence of the neutrinoo Scintillators can be classified as solid or liquid, organic or inorganic, and in terms of sensitivity to different types of radiation, ioeo, the fraction of the incident energy absorbed, efficiency in converting absorbed energy to useful light energy, and decay time of the emitted radiation, or the length of time after the radiation has been absorbed during which the substance emits 68% of the total light energy. There are many substances and combinations of substances which are scintillators, but only a relatively few have useful characteristics for nuclear radiation detection applications. Fortunately, there are some materials which have excellent properties. Some of the more common substances which are used here are anthracene crystal, sodium iodide crystal, zinc sulphide crystal and terphenyl. 8.26 Scintillation Mechanism For gamma detection, sodium iodide with a small amount of thallium as an impurity is often used. A simple explanation of the mechanism which causes the scintillation can be postulated. Quantum mechanics has shown that the electrons in a solid can be classified according to their energy level. The electrons are permitted to exist only in discrete energy intervals or bands. In a solid, most electrons are normally in the valence band. These electrons can not move freely throughout the crystal.. At a higher energy level, there is the conduction band of the solid~ Electrons in the conduction band are free to move about. The understanding of these bands has led to the explanation of a variety of physical phenomena associated with solids, e.g.,, the solid state detector to be described in a future lecture. In a perfect crystal, due to the discreteness of nature, electrons can exist either in the valence or in the conduction bands but not between. In an imperfect crystal and/or one contaminated with impurities in small amounts, such as NaI(Tl), discrete energy levels exist between these bands. The incident radiation imparts some of its energy to electrons in the valence band. These electrons can only go to the intermediate levels or the conduction band. Since the valence band represente a lower energy, more stable state, the excited electrons eventually return to the valence band and emit the difference in energy between the state they left and the state to which they have gone in the form of electromagnetic radiation. The only change of states which produce radiation appropriate for the photoelectric effect are between the impurity levels and valence band. These transitions occur in NaI(T1) with a decay time of 2~5 x 10-7 seconds. Since the incident particle loses its energy to the crystal in less than 10-6 seconds, 8 48

the electrons emitted from the photocathode can be considered as having been formed all at the same time, as the result of the one incident particle. Unless the count rate is greater than 106 per second there is little possibility of pulse pile up at the output of the photomultiplier tube. The crystal is transparent to the electromagnetic radiation produced since it does not have enough energy to lift an electron from the valence band back up to the same impurity level. In organic solids the idea is somewhat the same but no impurities are needed. The electrons are transferred to a higher vibrational energy state in which a part of the energy is dissipated by thermal excitation. Eventually a certain fraction of these electrons, in dropping down to the more stable vibrational energy states, will emit radiation with theproper wav - length for photoemission. The decay time for this possibility is 2.7 x 10-9 seconds for anthracene. The mechanism for luminescence in organic liquids is not well known. However, for inorganic gases such as xenon and helium, as the excited atoms or ions return to their ground states high frequency radiation in the ultraviolet is emitted with a decay time of about 10-9 seconds or less. Because no lower energy radiation is emitted relatively few photoelectrons are produced, but the fast decay time is cause for considerable research on this type scintillator. Probably the most easy to produce and obtain in quantity are the organic liquids, since it has been found that reagent organics are satisfactory scintillators. Experiment has shown that the amplitude of the photomultiplier output pulse is almost linear with the incident nuclear particle energy for energies of interest in this laboratory. Price gives the results of such experiments up to energies of 15 plus Mev. for a variety of particles (Fig. 7-4, 7-5). It has been found that ZnS(Ag) with a decay time of 10-5 seconds is excellent for alphas and many organics for O's. For a particular energy spectrum, one should check the literature to find the best scintillator for a given application. In mounting the scintillat'or on the photomultiplier tube it is obvious that all external light should be prevented from reaching the photocathode. Also, the interface between the scintillator and the photocathode is important because of the light which is reflected back into the scintillator because of the different indices of refraction of the two media. Light striking the interface making an angle, 0, with the normal to the surface will be totally reflected above the critical angle, G = sinl(n2/n!), where n, is the index of the scintillator. Air makes a very poor interface because its index is so low, giving a small critical angle. If it is inconvenient 8.49

to have the scintillator physically close to the photocathode, it has been shown practical to use light "pipes", sometimes as long as a foot or two, to serve as a guide for the light. There is always a distribution in the output of the photomultiplier tube due to the statistics of all the processes involved and the tube itself. The resolution of the output of the tube for a constant energy radiation particle impinging on the scintillator can be defined in any number of ways. Price uses the definition that the resolution is the ratio of the square of the mean voltage pulse amplitude to the difference of the mean of the square and the square of the mean. For gamma counting, the scintillator efficiency is defined as the fraction of y's absorbed in the scintillator. For a point source on the axis of the scintillator, the intrinsic efficiency is defined as the 1 / (1 - el ) dr where is the source solid angle subtended by the surface of the scintillator nearest the source, note x is the straight through distance the gamma travels inside the scintillator, For convenience the intrinsic efficiency is usually plotted as a function of y energy for a given shape of the scintillator volume with the source distance from its surface as a parameter. This efficiency is required if one desires to make an absolute determination of the source activity. The upper limit in counting is usually set by the resolving time of the scaler, about 2 x 10-6 seconds. To analyze a continuous distribution of incident energies or to investigate system resolution, it is convenient to use a channel analyzer. The analyzer is simply a piece of equipment which accepts only pulses between preset limits, ALE, when used as a differential analyzer and only accepts pulses above a preset limit, Ed, when used as an integral analyzer. On differential, the width of the voltage spread accepted is set by the "window" of the analyzer, AEo The differential analyzer gives the number of pulses in a given energy (voltage) interval, ZSE, above a base voltage denoted by Edial = Ed, which is continuously adjustable by the operator. 8 50

8.27 Analysis of Gamma Ray Pulse Distribution (Jointly with W. Smith). The units of Ed are not important, (for our equipment 0 < Ed 1000 units, 0 < 80 volts) as long as a relationship can be established to determine the energies of the gammas interacting in the scintillation crystal. A representative block diagram of the equipment is given in Figure 8.18. Details on the scintillation crystal and photomultplier are shown in Figures 8.19 and 8.20. Figure 8o21 is a typical well scintillation counter found in most detection laboratories. The abscissa of a differential curve is usually given in Ed units and the ordinate in counts per unit time per window. The pulses in /aE are initiated in the crystal by energy transferred to the crystal by electrons (details given below). The electrons are produced by the interaction of x and gamma rays with the crystal, These photons interact in three principle ways as indicated earlier. Consider the effect of each type of interaction on the observed differential spectrum. In a photoelectric absorption, the photon disappears and an electron of kinetic energy Ek= h~v w is emitted, w being the binding energy of the electron. The electron will lose all its energy in the crystal, and the X-ray emitted by the atoms returning to the ground state are soft, and therefore totally absorbed, Thus, the pulse will correspond precisely to the energy of the gamma ray. The pulses due to photoelectric effect will not be all equal, but will spread around the value Ey forming a peak, as shown in Figure 8-22. This distribution of the pulse size is da:e to many reasons, such as the random nature of the interactions in the crystal, the variation of the amount of light:reaching the photocathode, the nons-uniform response of the photosensitive saurface to unifom.size light pulses and the statistical variations in the phototube multiplication. In Compton scattering, part of the energy of the photon is transferred to an electron, and a secondary photon of energy + (1 COB O) mc2 8o52

8. 52 CRYSTAL INCIDENT i PHOTO- COUNT - 4-4 MULTIPLIER AMPLIFIER RADIATION Kl i\T*b INDICATOR HIGH EMITTED LIGHT QUANTA VOLTAGE SUPPLY Figure 8.18. Representative block diagram for scintillation counter. REFLECTOR PHOTOSENSITIVE LAYER CRYSTAL (No ) GAMMA RAY LIGHT ELECTRONS PHOTOMULTIPLIER TUBE TOTAL LIGHT TO TUBE NEARLY PROPORTIONAL TO GAMMA RAY ENERGY IF I ELECTRON EJECTS 5 FROM A DYNODE, II DYNODES RESULT IN 5" ABOUT 50 MILLION ELECTRONS OUTPUT Figure 8.19. Scintillation Counter.

8.53 MIA SHIELD INCIDENT FOCUSING FOCUSED TYPE UNFOCUSED TYPE (RCA 931-A) (EMI 5060) Figure 8-20. Detail on Two Types of Photo-multiplier Tubes. (Courtesy Tracerlab, Inc.) Well cap Well shield 2x2 Well crystal - Lead filled shield --— 8 16- 3/8" 6292 Photomultiplier Base Probe (cadmium case) '......1 - / 12"1* Base plate Figure 8.21. Typical well counter. (Courtesy RCL Inc.)

8.54 dN CPS dE AE 0.51 - - 0.51 POSITRON ANNIHILATION I K- 0.51 -_ ---- 0.51 - 0.5 --- BACK- I SCATTERING I j I PEAK- I NOISE/ PHOTOPEAK PEAK - PAIR P R ODUCTIO NCID ENCE II I I~~COMPTON I I EDGE I I Eb EC Ey Ed. H. E diol Figure 8-22. Idealized differential curve for rronoenergetic gammas

is ejected at an angle G. The Klein-Nishina formula describes how the probability of scattering varies with G, and indicates that more forward scattered electrons and back-scattered photons are produced than of any other energy. If we assume for one moment that all gamma interactions are Compton scattering and further, that the secondary photons always escape the crystal, the pulse size distribution would be as indicated, with a maximum number of electrons going in the forward direction. Actually, the scattered photons will suffer further interactions, and will be partially or totally absorbed, and the pulses resulting from Compton interactions will have a continuous distribution ranging from zero to the maximum energy that can be transferred to the electron. The energy of the backscattered photon for G = 1800 is Eb = ' 1+ Y mc2 The electron maximum energy is then: E Ec = E - Eb = YE 1 + mc 2E, which defines the so called Compton edge. It must be observed that in principle the photon scattered with energy Eb could suffer another interaction. However, if the values of EC and Eb are calculated for the current range of primary gamma energies, we get: E Eb Ec 0uo 0.169 0.169 0.331 1.0 0.203 0.797 5.0 0.242 4.756

showing that the range of variation of Eb is limited to a region where the photoelectric effect is p.redominant. Then, if -the backscattered photon is absorbed, the total primary energy would have been absorbed, and the pulse size will be under the photopeak. In most of the case, however, the secondary photon will escape the crystal, since it is directed outwards, and therefore very few counts will be noticed besween EC arnd the photopeak, giving origin to a characteristic valley in the differential curve, The spreading of the pulses around Ec is due to the same causes listed, above. Pair production can occur only when the eneergy of the primary ganma is above 1 02 Mev, The energy Ey of the photon is used to create the pair and to impart kinetic energy to the positron and the electron. This kinetic energy will be absorbed in the crystal, and finally the positron will annihilate with one electron of the medi=m. Two photons of 0. 51 Mev appear in opposite directions and three cases are possible: 1) if both photons are absorbed, then the total gamma energy E~ has been absortbed. in the phosphor, and the pulse will fall under the photopealk 2) if one of the photons escapes, the total energy deposited in the crystal is: Ea E7 0 51 and a peak will appear at that point, and 3) if 'both O 51 Mev photons escapes the crystal and the probability for that is considerably smaller, since they are ejected. in opposite dfirectiois the energy absoxrbed in the cryst'al is: Ea = ~ 1o 02 and a second and smaller pair productiorn peak will app(=ar superimposed to the Compton continuuom, It must be noticed that althboug.h sev!aral inte-rations can tak-e place successively,, the total ti;Le is short compared with the decay time of the crystal, so all of the~m will yield a single puzlseo Other peaks of intetre-st ae usually prestxit in the differentlal curves, although they are not due to direct interaction of the primary gamma with the crystal. If the scintillation detector is surrounded by a lead shield., many of the gammas will be back-scattered. towards the crystal, where by photoeffect they will produce a peak at energy Eb, Even wTTithout a shied., this peak could be noticeable if the crystal is close to any othter object, since back-scattering is predominant in Compton effect. If the source emits a posLriL, it will auihilatt.e after losing its kinetic energy, creating two photons of O51l Ma movingg in opposiite directions. One of 8.56

them will almost always reach the crystal and will show in the differential curve as a peak at O. 51 Mev. This photopeak is rather large, because the cross-section for photoelectric effect increases with decreasing energy. The second 0.51 Mev photon will or will not reach the crystal, depending upon the geometry (well type crystal, solid crystal) and the place where the annihilation occurs. If it does reach the crystal, a smaller peak will show at 1.02 Mev. It can also happen that one of the primary gammas, with energy E., will suffer a pair production interaction in the shield. The positron of this pair will annihilate also in the shield, but then only one of the two 0.51 Mev photons can hit the crystal, since the other one will go in the opposite direction. In that case, then, only a 0. 51 Mev peak will be observed. Finally, when we have two gammas of different energies, as for example in the case of Figure 8.22, the primary of energy E7 and the 0.51 Mev gamma from positron annihilation, it can happen that both will hit the phosphor at the same time, adding up their respective pulses to produce a coincidence peak at energy Ey + 0.51. o828 Details on the Conversion of Gamma Energy to a Measureable Pulse from the Phototube (Jointly with J. Trombka) The light emission spectrum for pure NaI is essentially the same as the absorption spectrum and is well below the visible range. To enhance recombination and to provide a "wave shifter" thallium impurity centers are added, which have the effect of providing intermediate energy levels in the forbidden band. The emission spectrum of Tl in Nal is lower than its absorption spectrum and provides a scintillation with wavelength in the visible region. This emission spectrum and the absolute conversion efficiency of the crystal ionization excitation into light emission is denoted by the conversion efficiency symbol Cn.p(X)2. The total light energy produced, starting with the initial excitation energy of the electron, Ee is 0 E - Et h Cnp(X) dXp Ee Cnp (1) and has the following spectral shape (fron Price). 8.57

4050Ao oi.016V,_NaI(Te) emission spectru.012 60 \ Photocathode "absorpti n" spectrum S() Cnp().oo8 _ 40 lgo/A~) a l \ \ j S(X) b( tt9) 1%/AO) watt.00 O _20 3500 3900 4300 4700 5100 X(A~) Figure 8.23. Conversion Efficiency C (X) and Spectral np Sensitivity S(X) of Type S-11 Photocathodes. The total area under this spectrum curve is roughly 550A~ (0. 016%/AO ) 8%. The conversion efficiency Cnp is dimensionless, 8.29 Conversion of Light Energy to Photo-electrons and Subsequent Multiplication If we place an end-window photocathode (SbCs3) surface in optical contact with the crystal, the fraction of photon energy which strikes this surface will be Ee (Tp * Fp) where Tp is the transparency of the crystal and of the phototube optical seal, and Fp is the non-escape probability. For a well reflected crystal both should be nearly unity. A fraction of this energy will be absorbed in the photocathode, giving rise to a number of low energy photoelectrons released inside the phototube. The efficiency of this conversion from photons to electrons depends on the absorption spectrum of SbCs3 and on the electronic stopping power of the photocathode and is measured by the sensitivity factor, S(X), shown also in Figure 8.23. S(X) is measured in units of amperes/watt and has a maximum of 0. 056 amps/watt at 4900Ao. This can be redefined in terms of the number of photoelectrons per ev of incident light 8.58

) q/dt /.amp dq yCoulombs dne yelectron) S(x) - _ | - r- dE/dt w dE - Joule dE \E ev Hence dne - S(), dne - S(X) dE dE From (1) above, dE dE2 = EeCnp(X) dx, so dne = EeCnp(X) S(X) d ~ Tp? Fp or the NUJMBER OF PHOTOELECTRONS EJECTED = ne and depends on the overlap of the spectra of Figure 8.23. 00 ne EJ Cnp() S(X) d. TpF (2) If we now define Fc as the fraction of such electrons striking the first dynode and if the subsequent phototube multiplication is M, the total charge produced at the phototube output, due to a single original y event is q = ne e' M~ Fc where e is electronic charge, or 00 q = Ee (TppFFc)M~ e SCnp(X) S(X) dX o If this charge collects on the input capacity of an amplifier without leakage losses, the resultant voltage pulse will have the maximum value o photopeak pulse from Csl37 (Ee = Ey = 0.661 M-ev) v(.66)( 5o4)'( 5 ) (o 6)(1- 6 o-Z 1 F9 ) Vin 2 l = 0o2 volts (4) 8.59

8.30 Pulse Height vs. Energy The interaction time for any of the three gamma interaction processes mentioned above is much shorter than the decay time of the scintillations in the crystal. Therefore, one gamma ray upon suffering an interaction or multiple interactions will produce only one pulse of photons. It is very important to observe that the number of photons in a given pulse will be proportional to the energy lost to the crystal as kinetic energy of electrons. Further, there is a linear relationship between the number of photons striking the photocathode and the electronic charge appearing at the output of the multiplier phototube. Thus charge will appear as a voltage pulse across the amplifier input capacitance. The pulse height in volts observed at this point will then be proportional to the ionization produced by the gamma ray in the crystal. Thus, the number of pulses per unit time at the phototube output is equal to the number of gamma photons interacting with the crystal per unit time; while the magnitude of each pulse is proportional to the ionization produced whithin the crystal by a given gamma ray. 8.31 Energy Resolution Finite Width of Photopeak The events producing pulses in the photopeak have a spread in apparent energy even though they represent full capture of the y energy (E = E ). yTotal e This is because there are statistical factors in the system which give a probability distribution in pulse height. Some of these factors are CORRECTABLE such as those associated with light collection optics. Every effort must be made to bring Tp, Fp as close to unity as possible. Some factors are INHERENT in the technique and set a severe bound to energy resolving power of the technique. If we define resolution as the full width at half maximum of the photopeak 2R 0- E1/2 E0 it will be found that the smallest value obtainable for the Cs peak is /v 6% (which for mediocre optics it may be 10 to 15%). This inherent limit is energy dependent and is due mainly to (a) statistical fluctuation in the number of electrons from the photocathode as given by Equation 2 and (b) to statistical fluctuation in phototube multiplication M. These two statistics must be "folded into" each other to determine the overall probability distribution. 8-60

The statistical variation in ne is truly Gaussian in nature and one can define the probability that an incoming particle of energy Eo will give a burst of photoelectrons of a number corresponding to some energy E1 as fE1 Eol2 ne(E1) a2 P e (6) 1 ne(Eo) The statistical naturjq of M is complex and not truly Gaussian. We will assume it to be approximately so, so that we can define 'the probability that the phototube multiplication of these ne(Ei)-will be of a magnitude corresponding to a pulse of energy E2 as 1E2 - Ei2 M(E2) b2?2 ()M(E) 7) The probability of obtaining a pulse height corresponding to an APPARENT ENERGY E2, originating from a pulse of true energy Eo will be the product of P1i P2 summed over all possible values of E1 (or E1 - Eo); the pulse heights will then have the value, from (3), (6), (7), q = C ne(E1)M (E2) dEl Cne(Eo) M (E1)| P1P2 dE1 Cnie(Eo) X P1 P dEt Letting e (E2 - E,), and x = (E1 Eo) D00 X2 2 q( C) = Cn / e e b e dX (8) Integration of this expression will lead to the distribution function 4 E2/2Cr 2 q(E) = Cne M e (9)

where 2a2 = a2 + b2; a2 is the dispersion (a = the standard deviation) of the distribution and a2, b2 are the dispersions of the photoelectrons' production rate and subsequent multiplication rate, respectively. It is shown in Nucleonics (10, 51 (1952) that the dispersion of the multiplication is a complex function of dynode multiplications, but photoelectron dispersion is just proportional to ne, the count rate: 2 = a2 ~rn - e ClEo 2 2 b2 am -- ne f(M) = C2Eo so a2 = C3Eo (10) Now from (5), (9), (10),R _ 261/2 Eo = ~/2 A.n 2. - (10a) E1/2 a (a) so R(E0) = 22 n 2 = 2.36a:~ Eo Eo and 1/2 1 (11) R(Eo) = 2.36(C3)/ () Hence a 10O resolution system for the Cs137 peak (E =.661Me ) should exhibit a value for the.41 Mev y from Au198 of 10%o (.661/.411)1/2 = 12. 7%. Integral of Photopeak +oo00 00 N = N(e) de N(E) e =2 — N(0) eF:~2 (12) _00 0 where N is the total number of counts under the photopeak and N(O) is the maximumPcount per unit width. 8062

8.32 Detection Efficiencies We now look at the problem of the detection efficiencies. In the following discussion we will not consider pair production. The analysis of course can be extended to include this process, The cross sections of interest therefore will be those for photoelectric absorption and for Compton scattering. Consider the case of amonoenergetic point source. We assume that there is no scattering from surrounding source materials Source Io NaI(Te) Crystal Figure 8.24. Typical Source-Crystal Geometry. If Io gamma rays per unit time are emitted from the source, then Iio/4A will be the number of gamma rays incident upon the top surface of the detector (where d _ 2t sin Qd~ and by integration for G = 0 to 0 = QmQ = 2rt(l - cos Gm). Define the total intrinsic efficiency, GTi as the probability per unit solid angle that the gamma ray suffers a first collision, then NT = Io Ti (13) where NT is the number of gamma rays emitted from the source which suffers first collision in the crystal, Since one pulse will appear for every primary collision, and since only the pulse height is effected by the type or number of interactions for a given gamma ray, the number of pulses per unit time detected, independent of pulse height, will equal NT. Therefore, if we add up all the pulses produced per unit time under the pulse height spectrum, we obtain the number of primary interactions occurring per unit time. If AT be the total area under the pulse height spectrum, then A = NT = Io -Ti (14) 8.62c

If we consider any interaction in which a gamma loses energy, and this energy eventually is converted to scintillations by absorption, then t can be defined as the linear absorption cross section for first interactions due to photoelectric and Compton interactions. Then with respect to Figure 8.24, e'UP will be the non-interaction probability along the path p within the crystal, and (1 - e~'P) will be the probability of suffering a first interaction along p. Then m -11) d2_ Ta (1 - e d)) - Total absolute (15) o efficiency is the probability that if a gamma ray is emitted from the source, it will interact in the crystal, The integration is carried out over the solid angle subtended by the source and top of the crystal. Then we define CTi -Ta Total intrinsic efficiency (16) 4c which is the fraction of those gamma incident on the crystal face which interact with the crystal. Both integrations are carried out over the solid angle described above. This then is the efficiency ETi shown in Equation (13). This efficiency has been studied as a function of source detector distances for a number of energies and for several crystal sizes. In terms of the analysis, it is usually simpler to study the area under the photopeak only. This area can be determined much more precisely than the total area. There are two major reasons for the difficulty in obtaining the total area, First, it is rather difficult to eliminate all scattering effects due to the surrounding materials. These will appear as pulses in the Compton continuum. Secondly, pulse height analyzers cannot detect all pulses down to zero pulse height, for below certain pulse height levels, the equipment noise and the thermal noise of the phototube completely interfere with the detection. We, therefore, consider AP - the area under the photopeak Epi - the intrinsic peak efficiency 8, 6

or, ECi is the fraction of those gammas striking the crystal face which are totally absorbed. If both sides of (14) are divided by - we obtain A = tI0 1 Epi where T;Ti = PT(Ti17) PT is the peak to total ratio Ap/AT. The total number of absorptions in the photopeak may be obtained from Equation 12. (When using a pulse height analyzer to measure these pulse height spectra - histograms rather than true differential spectra.are obtained because the analyzer measures all pulses in a finite.iE increment about E. Therefore N(o) of Equation 12 will equal Amax/tE where Amax is the experimentally determined maximum of the photopeak in the measured pulse height distribution. Thus P' = AN -Amax ALE where a and LE are measured in the same units (e.gy, either pulse height or energy) and a is determined from Equation lOa.) Then, using the known geometry of the source-crystal assembly and the tabulated values of efficiencies, that fraction of the gammas emitted by the source which is totally absorbed in the crystal and appears in the photopeak is obtained, Hence, the source strength may be absolutely daetermined. 8 33 Experiment Using a Scintillation Counter Please use extreme caution in working with photomultiplier tubes so that they will not be damaged by excessive voltage, by exposure to light while high voltage is applied or by breakage. Use the G.Eo scintillation counter for parts 1 through 6, and the RCL well type counter for rest of the experiment. 1) Obtain the count rate versus voltage curve with no source and no scintillator in order to find out where tube noise begins to produce counts and to check the light tightness of the detector. 8,65

2) Repeat the procedure with the ZnS screen in place, but no source. 3) Obtain the count rate versus voltage curve using thorium foil and the ZnS screen. 4) Show that you were counting alpha particles by covering the thorium with a suitable absorber. The thorium, together with its disintegration products, emits betas as well as alphas. The efficiency of ZnS in converting ionization enery into light is the same for alphas and betas, With these points in mind, explain why you were counting primarily alphas with the ZnS scintillator, 5) Obtain the count rate versus voltage curve with the anthracene crystal and no sources 6) Repeat the procedure using the bismuth 210 (RaE) source, and then the carbon 14 source. Compare these two curves, and explain why they do not have the same shape. 7) Obtain the gamma background count rate versus voltage curve with the well type counter, 8) Obtain the characteristic curves for cobalt 60, cesium 137, and cobalt 57. Comment on the similarities and explain the differences. 9) Decide on an operating voltage for the cobalt 60 source. At this voltage, count the background and the source as the scaler input sensitivity is varied. Calculate the ratio of the background count rate to the count rate due to cobalt at each sensitivity setting. Explain why this ratio may vary with sensitivity. 8.34 Experiment on Scintillation Spectrometry Use a scintillation counter with a single channel pulse height analyzer in this experiment, Please note that the super stable high voltage supply can deliver a lethal electric shock, It is recommended that you use 1000 volts for thep hotomultiplier tube. The tube may require a positive or a negative high voltage, depending on whether the photocatode or the collector anode is held at ground potential. It is recommended that you use an amplifier gain value such that the cesium 137 peak occurs at an E dial setting of about 350. 8066

1) Both the E and AE dials of the single channel analyzer are divided into one thousand divisions. One division on the E dial, however, corresponds to many divisions on the AE dial; that is, the graduation of the AE dial is much finer than that of the E dial. Determine experimentally what percentage of the maximum E dial setting is equivalent to the maximum AE dial setting. This figure is called the maximum window width, For complete checking of the aE dial, one must ensure that the window width is not dependent on the E dial setting, and that the LE dial is linear. These checks can be performed simply by using a pulse generator. 2) The next step is the calibration of the E dial; that is, the determination of the proportionality factor between E dial setting and energy. With an approximately 1% window width, locate the full-energy peaks of cobalt 60, cesium 137, and cobalt 57. Draw the E dial calibration curve, and decide whether it is linear and whether zero setting corresponds to zero pulse height; if not, please call this to the attention of the instructor, 3) Determine the complete pulse height distribution curves for cobalt 60, sodium 22, cesium 137, and cobalt 57. Please note that you need the background pulse height distribution 'to do this. Use a 1% window, and take data at as many E dial settings as necessary to determine the shapes of the curves. When drawing the curves, graduate the horizontal scale in both E dial and energy units. Point out and explain the characteristics of the obtained curves. Identify all peaks; give two reasons why the full energy peak corresponding to the lower energy gamma of cobalt 60 is higher than that of the higher energy gamma, in spite of the fact that equal numbers of these two gamma are emitted, Compare the sodium 22 spectrum obtained with an unshielded solid crystal with that obtained with a shielded well type crystal. Point out and explain the differences in the locations and the relative magnitudes of the peaks. Determine the energy resolution of your equipment for cesium 137 and cobalt 57 gammas in terms of the absolute width and the percentage full width of the photopeak at half of the maximum value. Explain the reason for the difference in resolution associated with the two isotopes. Please describe how you would determine the activity (in gammas emitted per unit time) of a source, given that the intrinsic peak efficiency of your crystal for the particular geometry used and the particular gamma energy of the source is g.,

8. 35 Neutron Detection Because neutrons are unacharged, their passage through matter does not create ion pairs. For our purposes, any neutron induced reaction which has either charged particles or gamma rays as a reaction product is potentially useful. The cross sections for neutron reactions are generally energy dependent. By measuring the effects of the products of the neutron interaction, one can not infer both the neutron flux and the neutron energies, since they sxe independent. In most measurements there are a spectrum of neutron energies. Energy bands can be selected from this spectrum mechanically using time of flight, crystal spectrometry or absorption techniques. Or, if the spectrum is known, egg.g, Maxwell'Boltzmanr, the count rate can be related to the incident flux. The most common gaseous chambers employ boron 10 or fissionable material. The solid boron or fissionable material is coated on the inner walls of the chamber, In the thermal neutron-boron reaction an alpha is produced and charged fission products as well as lighter particles in the neutron-fissionable material reaction. By proper design, these charged particles produce ion pairs in a convenient chamber gas. Since boron exists in the gas phase in the form of BF3, the chamber can be filled with this gas to accomplish the desired result of neutron detection. Because of the high specific ionization associated with as and the massive fission products, these chambers can be used easily for pulse detection. Boron lined chambers usually have a plateau with a greater slope than BF3 tubes because of the variation in the a energy lost in the solid as opposed to most of the a energy being dissipated in the BF3 gas. Fission chambers have the greatest slope in the count. rate vs. discriminator voltage with chamber voltage as a parameter, because of the large fluctuations in primary ionization produced by the fission fragments, whose range is very small. The fission cha;mber must be operated as an ionization or proportional chamber to discriminate between the fission fragments and the background of a,, and y. Fast neutrons are usually measured through the intermediate step of an (n,p) scattering reaction, i.ex, the neutron scatters with either a hydrogen or a hydrogen containing molecule and an. energet-ic recoil proton is produced. The ionization produced by the proton is measured4 A very common method of measuring neutron flux1a levels in a reactor, when a minimum flux perturbation is required., is the use of small pieces of metallic foil. The neutrons produce reaction products in the foil which are radioactive. When the foil has become sufficiently radioactive to give good counting statistics, the decay radiation is measured by standard means. If the energy distribution of the neutrons striking the foil is known the flux corresponding to the measured foil count rate can be calculated, It is clear that this method can only givre 08 68

integrated neutron flux levels. Gold and idium are common foils for flux measurements. By enclosing the foil in a material with a large resonance cross section over a certain energy interval it is possible to restrict the energy spread of the neutrons entering the foil. Cadmium, which has negligible absorption for energies greater than about 0 5 ev and a very large cross section at lower energies, is often used to shield the foil from thermal neutrons. By measuring the foil activation with and without the cadmium shield, one can get a direct measure of the thermal neutron flux. The often quoted "cadmium ratio" simply refers to the ratio of the activation without to that with the cadmium present. The cadmium ratio is greater than or equal to one in a "thermal" reactor. A little thought should give the characteristics which an ideal foil would have with regard to cross section, type and energy of radioactivity and decay time. Dalton, at this university, has recently written a thesis on the flux depression inside thin foils which gives theoretical predictions of the neutron flux in excellent agreement with experiment. Scintillation techniques can be used for neutron detection by utilizing the products of a neutron reaction to form ion pairs in the scintillation material. In high density scintillators, compensation for induced background pulses is often difficult. There are a host of development problems. New scintillators and designs are being worked on at this time. Other meas of measuring the products of a neutron interaction can be envisioned and are used in one form or another. For example, the recoil proton path can be captured on a photographic type emulsion, or the heat of reaction can be measured by the thermoelectric effect, or the amount of evolved gas can be measured, etc,, etc. In working with neutron sources it is important to keep personnel dose rates below the permissible levels, Figure 8.25 gives the maximum permissible neutron fluxes as a function of neutron energies. 8.36 Experiment on Neutron Detection Using BF3 Tubes Use a BF3 detector for this experiment. In order to avoid electrical shock, please turn off the high voltage when handling the detector, The main objects of this experiment are the determination of an operating point, in terms of amplifier gain, discriminator setting, and high voltage, at which neutrons can be counted in the presence of gamma radiation; and the measurement of neutron fluxes, Use the cathode ray oscilloscope throughout this experiment. 8~ 69

Basis: 14 x 106 neutrons per square centimeter equivalent to a dose of 1 rem. (See Federal Register Title 10, Chapter 1, Part 20: Standards for Protection Against Radiation. ) UJ 10 / - _ (. 7 w CC I M 4, _ I C. Pu-BeSOURCE., WI 0 DISTANCE FROM POINT SOURCE NO SHIELDING Prepared by,:: g,,,~ ~ ~ ~ ~ ~ ~ ~ ~~Peae / GL. Gyorey~~~~~~~~~~~~~~~~~~~~~

1) If the gain is too high. the amplifier will be overloaded by the larger pulses and will be rendered inoperative for a short time. The optimum gain is the one for which -only a few counts are registered at a discriminator setting of 1001 To find this value, set the gain to its maximum value, use the high voltage setting recommended by the manufacturer of the detector, and start counting neutrons. Reduce the gain until the desired value is reached, Ask the instructor to remove the neutron source. 2) Using the gain setting just determined, turn off the high voltage and set the discriminator to its lower limits. The counts should now be due to noise, Determine the discriminator setting at which the count rate due to noise becomes negligible1 3) Turn on the high voltage and expose your detector to a gamma field of a few mr/hr. The.count rate should now be due to the gamma radiation. Determine the count rate versus discriminator setting characteristics of your apparatus for gamma radiation, Remove the gamma source, 4) Expose the detector to radiation from a source emitting mainly neutrons, placing both the source 'and the detector into a block of paraffin. Obtain the count rate versus discriminator setting characteristics, and compare it to the one 'obtained for gamma radiation, Decide what discriminator settings you would use for detecting neutrons in the presence of gamma radiation and for counting neutrnons when no appreciable gamma field. is present. If the two values are not the same, use the latter for the rest of the experiment. 5) Obtain the count rate versus voltage curve, and determine wheather the high voltage value that has been used is a desirable one, If not, please bring this to the attention of the instructor. 6) Determine what fraction of your counting rate is due to thermal neutrons by taking counts with and without a cadmium shield, which absorbs practically all of the thermal neutrons around the detector., From your count rates in the paraffin block and outside the paraffin shield, calculate the thermal neutron fluxes at these two points. Calculate also the radiation dose rates at these points due to thermal neutrons, given that exposure to a thermal neutron flux of 670 neutrons/sec/cm2 is equivalent to 2.5 mrem/hr. 8. 37 Thermal Neutron Flux Measurements in a Nuclear Reactor Using Activation Techniques The activation method of neutron flux measurement makes use of the production of radioactive nuclei by the absorption of neutrons in a detecting material. The radioactive nuclei so produced decay with the emission of nuclear

radiation and can thus be detected. The most widely used detecting materials are indium and gold in the form of foils or wires. The rate of production, R, of the radioactive nuclei depends on the absorption cross section of the detecting material and on the neutron flux in the followinrg manner: 00 R ND a (E) 0 (E) dE where ND is the number of nuclei in the detector, a the absorption cross-section, 0 the neutron flux, and E the neutron energy. A rough plot of ~a for low energies (in general) is given below in Figure 8.26 -'1 1 1~\ In: 1.44 ev I:I /!\ERES for Au: 5 ev ar IE Co: 120 ev ET....c c. R SaI ET ECC ERES E Figure 8.26 Typical Neutron Absorption Cross Section ~a vs. Energy, E. For the FNR - Maxwell-Boltzmann Distribution. ET 0.03 ev rETC 2. ev E t ET ETC ECC E ~ 20 mev Figure 8.27 Neutron Flux, 0,v,x, Energy, E Here ET is the most probable energy of the thermal neutrons, ETC is the thermal cutoff energy, and ECC will be discussed below. 8 72

We may now write 00 R ND MaET th+ T Ca (E) (E) dj ETC where CM is a factor which is applied to aET so that CM aET gives the average cross section for the thermal flux. In order to measure Oth' one must somehow differentiate between the quantities of radioactive nuclei produced by slow and fast neutrons. This can be done by first making measurements with a cover placed over the detecting material. This will screen out one of the two neutron groups. A material for which a varies as shown below is needed. Very High x- Very' Low ECC E Figure 8 28 Idealized Absorption Coefficient vs. Energy for Screening Slow from Fast Neutronso Here ECC is the cutoff energy. We would like to h.ave ECC k ETC. A rough plot of aa for cadmium is shown below 7200 r a a (Barns) Z80 0 - Effective ECC 4 ev 300 cc I,........,..- *. w. _ _.t _.176 0 ev Figure 8~29 Approximate Absorption Coefficient for Catdmium Now if we assume that no neatrons with E < ECC get thirough the cover, and that all neutrons with E > FCC go thro gbh it' we have'EC:=% (F dE ECC With a bare dete'dtor, that is, with no cover, we have ECC 1b[C PDET a th+ aa (E) (E)dE+ * a (E) 0 (E)dE] ETC ECC

Rb/Rc is called the cadmium ratio. Taking the difference of the two rates: Ecc CC Rb Rc NDL CM aETth+ /E aa (E) (E) dE] ETC Let us now look at the integral term. For the energy ETC < E < ECC let aa;(E) a J/ 4E and 0(E) 0e/E, where aO and 00 are appropriate constants. We now may write: ECC ECC -3/2 E E ETC ETC a0 o [ol E/]CC 2a0 L.L. TC TC CC Buta - th so thata = f025th th = ~ t where ath is the value of a~j(E) at.025 ev. J, 025 Also, aET = Eth so that ETC CC If we define 5th tCM aE = /.025 th T VET Rb- Rc ND ath Oth + ET athT0 (CC0 So that finally we may write Rb - Rc = ND %th (0th + k 00) where k is a constant. For example, for ET =.03 ev, ETC =.2 ev, ECC =.4 ev. CM = J:/4 8.74

Correction Factorsl, 1/. 845 for.75" dia, 5 mil In. 1/. 84 for 1, 5" dia. 5 mil In 1 /o86 for 1. 5" dia, 5 mil Auo REFERENCES4 1, Deutsch, R. W, "Computing Three-Group Constants for Neutron Diffusion," Nucleonics, Vol. 15, No. 1 (1957). 2. Dayton, I. Eg, and Pettus, W, G., "Effective Cadmium Cutoff Energy," Nucleonics, Vol. 15, No, 12 (1957),. 5. Tittle, C. W,, "Slow Neutron Detection by Foils" I- and II, Nucleonics, Vol. 8, No, 6 and Vol. 9, No. 1 (1951). 4, Martin, D. H., "Correction Factors for CdaCovered Foil Measurements,' Nucleonics, Volo 13, No, 3 (1955). 5.o Gallagher, T L, "Foil Depression Factors for Indium Discs Detectors.," Nuclear Science and Engineering, Vol. 3, No, 1 (1958), 6. Klema, E. D., and Ritchie, R. H. "'Thermal Neutron Flux Measurements in Graphite Using Gold and Indium Foils," Phys" Rev,, 87, 167 (1952)a 7. J. L. Shapiro et al, '"Initial Calibration of the Ford Nuclear Reactor," MMPP 1O10-1, April, 1958, 8,38 Experiment on Neutron Detection by Induced Activity You will be given a number of foils which have been exposed to neutrons in a subcritical nuclear reactor assembly. You will also be given the following data:.the time of insertion and withdrawal of the foils, the weights of the foils, and their relative positions during irradiation, By measuring the activity of the foils and makling use of the supplied data compute the relative magnitudes of the neutron flux at the points where the foils were exposed, and plot the logarithm of some quantity directly proportional to the neutron flux against position., Please state clearly what quantity you are plotting. The plot should show the standard deviations associated with the quantities which are plotted. 8Q75

k =.258 Now if ~th >> k sO, and it usually is- we may take Rb - Rc a; ND %th Oth One should, however, distinguish between the thermal flux and the subcadmium flux at least in principle. Reference/l-discusses the determination of ET and ETC, and reference 2 the determination of the effective ECC. Up to this point we have dealt with the rates of the production of the radioactive nuclei. Now we must relate this quantity to one that we can measure. The quantity that we measure is the rate of disintegration of the radiocative nuclei at the time of measurement, which is some time after the removal of the detector from the neutron flux. The buildup of the radioactive nuclei during irradiation is described by the equation dN R R X NR dt If NR (t = 0) = 0: R (1- e ) NR (l-e ) Here t is the time of irradiation, R the rate of production of the radioactive nuclei, NR the number of such nuclei, and X their decay constant. The rate of disintegration A during irradiation is A =NR R R(1 e ) As t- oo, A - R, the so called saturated activity A,. A = R -Xt 1 - e After removing the detector from the neutron flux at t = tl, A will fall exponentially: 8.76

A-X(t —t)t -Xt A(tl)e =A(tl) e e The time behavior of A is sketched below: - - ~ - - - - - -1- - C- - - — I~ - -- - R = AA(t) irradiate I1 count 3 Figure 8.30 Detector Foil Activity During Irradiation and Counting. Counting the decaying nuclei with counting efficiency f from t2 to t3, the number of counts, C, is: t3 Xt -.xt C f A(t ) e:e dt + background t2 = fA(t!) et K e + background fA(t1) eXtl 1 (e Xt2 e Xt3) + background tl but A(tl) R(l-e t), and if we define Ccorrected C background, then R e t1 Xt2 -e t3 Ccorrected X1 so that Ccorr/f (eXtl)(et2 e 3) The analysis up to this point has neglected the following facts: 8 yy

1) Some neutrons with E < ECC pass through the Cd cover. 2) The Cd cover stops some of the neutrons with E > ECC. 3) The presence of the dector with its relatively high absorption cross section depresses the neutron flux. 4) The detector has a finite thickness, and therefore the inner volume elements are shielded by the outer ones, that is, the activity per unit volume is uniform. A detailed discussion of the above factors is beyond the scope of this paper. Only an indication of the results of some experimental investigat. ons will be shown. References 3 through 6 should be consuled for further information. The first two effects mentioned above can be investigated by irradiating detectors with different thicknesses of Cd cover. A typical plot of AX vs. Cd cover thickness. Correction Factors: \ 1.1 for 5 mil In, 40 mil Cd " 1.02 for 5 mil Au, 40 mil Cd 00 0 10 20 30 -40 Figure 8.31 Detector Foil Activity vs. Cadmium Cover Thickness. The shape of the curve depends on several factors, such as the material and dimensions of the detector, the medium in which the measurements are made, and the Cd ratio. The steep slope at and below 10 mils of Cd thickness shows that more than a negligible number of sub-Cd neutrons reach the detector. More than about 20 mils of seems to reduce this leakage enought so that only a small slope remains indicating the absorption of the epi-Cd neutrons by simply extrapolating the slope to zero Cd cover thickness. Effects 3 and 4 can be investigated by irradiating detectors of different thicknesses, plotting the activity per unit thickness against detector thickness, d, and again extrapolating to zero thickness. The slope of such a curve will depend on the detector material and dimensions, and the medium in which the measurements are made. A typical plot is shown below 1. 0 A./d..9 1 2 3 4 5 6 (mils) Figure 8.32 Detector Activity, Per Unit Thickness vs. Detector Thickness. 8~78

8 39 Fundamentals of Junction-Type Solid State Ionization Detector (Jointly with G, Brown) Rapid progress has been made in solid state detector technology since McKayl first measured alpha particles with a pan junction in 19514 Detectors are now available which are suitable for the measurement of charged particles and gamma radiation. Although the principles of operation are well known to semiconductor physicists, many workers have not had the opportunity to obtain more than just a descriptive understanding of the solid state detectors. The relations obtained in this section for barrier layer-depth. and electrostatic potential, as a function of reverse bias, apply to both the p.-n junction and surface barrier detectors. In addition to specific references in this paper, references 4 to 7 should provide -helpful information' for further study. Conventional notation2 is used wherever possible. Because this- detector is not considered by Price, considerable mathematical detail will be given, 8,o 40 Junction Detector Consider a crystal of a group IV element, such as silicon, with a very small amount of group III impurity such as boron homogeneously distributed in the lattice. The group III atoms act as electron acceptors in order to satisfy the bond requirements of the surrounding group EIV lattice elements. The positions in the crystal from which these electrons migrate are positively charged relative to the negatively charged group III lattice position, Since the group III atoms are fixed in the lattice, most of an electric current would be carried by positively charged '"holes"', moving through the crystal When the majority of the charge carriers are holes, the crystal is p:-type. If the impurity had been group V, such as arsenic, an electron would be available to the lattice from this donor impurity9 because of the low dissociation energy, resulting in a positively charged ion in the lattice and most of the current carried by negative electrons, When the majority of the charge carriers are electrons, the crystal is n-type. The silicon p-n junction detector is made by diffusing a group V element into a silicon crystal with a group III impurity. The transition region in the c:rystal is described by the distribution of the equilibrium concentrations of electrons, holes and ionized impurities- The n and p type portions of the crystal have low resistivity compared to the transition region. When an external voltage is applied, essentially all the potential difference takes place across the transition region. Since the width of this region is usually less than 10"cm, high electric fields are developed When an ionicing particle enters this region and produces electron hole pairs, i,e., raises electrons from the valence band to the conduction band, the electrons and holes are swept across the barrier in opposite directions by the electric field, The resulting induced charge gives a measurable voltage pulse in the exernal circuit, 8479

8, 41 Fermi Statistics The probability, f(E), that a quantum state of energy E is occupied by an electron at equilibrium is, (E-EF)IkT f(E) = l/[e + 1] (1) 2 and is known as the Fermi distribution function. The Fermi energy level, EF, can be given physical significance as the half occupancy energy since f(E) < 0.5 for E > EF. (1) is valid for particles which obey the Pauli principle. The probability of non-occupancy of a quantum state is l-f(E), or equivalently, is the probability of occupancy by a hole. The concentration of allowed energy levels in the conduction band which lie in the energy interval dE about E is given by Dc(E) dE = ic EdE (2) where Ec is the electron energy at the lower edge of the conduction band and Nc is the effective density of states in the conduction band. Explicitly, N = 2 (+2(3) n2 The concentration of electrons in the conduction band is given by 00 n = 2 Dc(E) f(E) dE (4) Ec The factor of 2 accounts for the spin degeneracy for each energy level Assuming that EF < IEc - 3kTI gives f(E) A exp [(EF - E)/kT]. The solution of (4) is n N exp [(EF -Ec)/kT] (5) The concentration of allowed energy levels in the valence band in dE about E is analogously given by 8.80

kT kT w 2(rmpkT ( 3/2 d where Nv = 2( ----)- / is the effective density of states in the valence band, and mp is the effective mass of the hole at the uppermost energy of the valence band, Ev, Similarly to (4), the concentration of holes in the valence band is given by p= 2 DV(E) [ -f(E)J E with the solution p ' N~ exp [(Ev-EF) kT] (7) For an intrinsic semiconductor, n-ni =n p = Pi Equating (5) and (7) and solving for the Fermi level in an intrinsic semiconductor, (Ev+Ec) + kT n (8) 2 4 mn At thermal equilibrium, np - n'.. Thus, from (5) and (7), ni fi eap [-(Ec-Ev)/2kT] (9) It is shown2 that f(ED), the probability that a donor energy level be occupied by an electron, is 1 f(ED) 1= /[ - exp (EE F)/kT} + 1] where ED is the electron energy at the donor level. If the concentration ofdonors is nD at ED, nDf(ED) are neutral and nD[l, f(ED)] are positively charged and a corresponding number of electrons are raised to to the conduction band Fo the condition (ED - 3kT) > EFo nD ~ nD+, i. eF, the impurities are totally ionized. 8 81

For an acceptor impurity, f(EA) = 1/[ 2 exp {(EA-EF)/kT} + 1] (10) If nA is the total acceptor concentration, nAf(EA) A_ will be occupied by an electron from the valence band, leaving a corresponding concentration of holes in the valence band. For (EA + 3kT) < EF, nA ~ nA. 8.42 Effect of Impurity Concentration Change and. Bias on the Fermi Level In a region where the space charge d.ensity, p, is zero, the condition of electric neutrality is given by nD+ nA = n- p nD - nA (11) Substituting for n and p from (5) and (7), nD - nA = N exp (E-Ec)/kT] N exp [(E-EF)/kT] (12) When nD >> nA, the Fermi level on the donor side of the junction can be found directly from (12), EF = Ec - kT Yn (Nc/nD) (13) When nD<< nA, on the acceptor side, EF - Ev + kT en (Nv/nA) (14) When nD = nA' (Ev+mEc) kT EF = + kT n (N/Nc) (15) 2 2 Note that (15) is almost identical with (8) which described EF in an intrinsic semiconductor. Since Nv and Nc are on the order of 109cm3 at 300~K and the impurities are on the order of 1021cn13 and less, it is possible for EF < Ec Consider the effect of an externally applied. potential or bias, VOX Using a reverse bias, i.e., the n side of the junction is made more positive with respect to the p side, the Fermi level is displaced by an amount eVo, 8.82

since thermal equilibrium has been destroyed, As a first approximation, it can be assumed that the spatial forms of n and p are not changed, in other words, that the applied voltage can be superimposed on the internally developed. space charge potential, 8, 43 Depletion D epth and Barrier Capacitance The ratio of the depletion depth to the range of an ionizing particle gives an estimate of the fraction of its energy transferred to the crystal. Since linearity between incident particle energy and detector pulse output is desirable, the depletion 6-depth foor a given detector should be known, By relating the internal detector potential -to this depth,b theoretically, the depth can be found by a simple measurement. Alternatively, if a known energy is deposited, and the output pulse observed., the depth. can be found from the capacitance per unrit area, Consider a pn juncution geometry for which the space charge is zero on the p side at x < p nD+= nA at x = 0 and 'the space charge is zero on the n side at x > xn. In. generai, n + will depend on the method used to make the diffused junction and the degree of perfection of the crystal. Ideally, nD(x) nD(xp) erfc (x/2 D{)~, wTrere D is the diffusion coefficient for impurity atoms in the crystal at soame chosen temperature for a diffusion time, tt, This distribut-ion falls off sharply with x and makes an analytic solution of the potential problem unattainable., The junction is defined at the position in the crystal wnere npD+ nAO The Fermi level is constant, however, throughout' tthe entire crystal when no external bias is applied., From (5)i and (7 the distributions. or electrons and holes on eithe.r side of the junetion are know.a, nn(x) c exp [ (EFe E (x)x/kS] (16) 5:x: =. E M Ep) / kT S np X) exp ((B (X k",7 -Since n to nD for x > ad ~9 nD >> ni aud p.: nA for x < - nA >> n, i e., outside the deple'tion region, n(x > xn) n: exp.(EB =)/kT] (17) and 8,83

Substituting for Nc and Nv from (17) into (16) and replacing Ecn(x) and Ecp(x) by Ec(x), and Evn(x) and Evp(x) by Ev(x) gives, n(x) = nD exp [(Ecn-Ec(x))/kT] (1s8) p(x) = nA exp [(Ec(x)-Ecp)/kT] since Ev(x) - Evp:= Ec(x) - Ec The crystal energy levels can be replaced by the electrostatic potential, since E = -eV giving, n(x) = nD exp [e(V(x) - VD)/kT] (19) p(x) = nA exp [-eV(x)/kT] where V(x < -xp) = 0 and V(x > xn) = VD. The potential drop across the depletion region is found from (19) by solving the ratio: p(x < -xp)/p(x > xp) P= n/pp p exp [eVD/kT] f or VD. Using pp nA,' P ni2/nD gives VD T n (nAnD/n) (20) e The space charge density is given in general by p(x) = e[p(x) - n(x) + nDr(x) - nA(x)] (21) Assuming the permittivity of the junction region is constant, the one dimensional electrostatic potential is obtained by solving the Poisson equation, = - p(x)/e (22) dx2 -For convenience, assume nD+(x) is a constant, nD+ for 0 < X < xn, and more reasonably that nA_(x) is constant for -= < x < 0, and nD+ = nI- = 0 otherwise. Combining (19), (21) and (22), dLD2vne kT n.1 n - (23) where nD+ = 0 for x < 0, and nA_ O0 for x > 0. Multiplying (23) by dV/dx and integrating over x, gives (~) 2enD - kT(e ( 1- ee]V"VD)/kT> \ (24) 8.84

For O < x < x, and n (dY _ 2enA- [kT _ 1 + e V(x) (25) For -Xp < x < 0. Equating (24) and (25) at x 0 and rearranging gives, V(O) [nD+ - nA] kT [nD+ exp {e(V(O) - VD)/kT} (26) + (nA -)exp {- eV(O)/kT} = nD+ VD - _ [mD+ + nA-] e which is useful for determining V(0). The electric field, Ex -, is known for the entire depletion region 'dx from (24) and (25).. However, because (24) and (25) are non-linear, no closed solution for V(x) has been found. Making a further approximation3, let p(x) = e[nA - nD+] (27) where nA. = 0 for x > 0 and nD+ = 0 for x < 0. The solution of (27) proceeds straightforwardly. Using the boundary conditions that dV dV V(-xp) 0=, 0, V(Xn) VD and O dx Xn gives enA enA EV(x) = - (x + Xp 2x~ and E(x) enD+ (xn x) (x = VD en-D+ (xn _. 82 2c 8,85

The maximum field, at x 0, is E enAxp = "enDxn (29) max se 4 and, also, from (28) at x 0 VD = e (nD 2 + nAXp) (0) The condition of electric neutrality is, nAxp =nDxn, i.e., 00 -00 Solving (30) and (31) for the depletion widths, nD > nA, Xp, (.2E:VD/enA)l/2 -,(2EVD~IJ~~, 2 1/2 (32) X, _' (2cVDnA/enD ) The barrier capacity per unit area, C is given by C = _:(enA/2 /2 ()33) X + Xn Xp Another impurity distribution which partially fit:s:.mpiria:1; resuit&. i:n the g;raded-' junction, p(x) = a x where a is a -= nD(Xn) nA(xn) = nD(-p) - nA(-Xp) Xn -Xp The solutions are a (x2 xp2) V(x) ea) x 2 2c 3 8.86

ea 2 22 and, Ex = 2 x - x ~~~Xn. 0 x Xn giving x = xn.From (35) 1/5 and,i a their usefulnessu 8&44 Balance Equation and Pulse Output The equation which describes the position of the electrons and holes produced by ionizing radiation in the depletion region is common to many branches of science2, i giving.,= xnFrom (9) and~~~~~at J is the total current, J = eE [n + p] + e L[Dn DpVp] and tn is the electron mobility, Dn is the diffusion coefficient for electrons, etches of science 6&87 etc.~ ~ ~ ~ ~ ~ ~ ~~~~~~~~ n ~~~Ln n~~88

The approximate location and time of creation:of ion pairs can be estimated, Depending on the degree of approximation, (39) can b.e solved. either in closed form or by machine it.eritive procedures. The charge induced on the electrodes by the -electrons and holes is known. The voltage pulse form to the scaler or analyzer can be calculated as a simple RC circuit. To get an appreciation of the magnitudes of the various quantities associated with these detectors, consider a particular p-n silicon junction detector. 8.45 Range of a Particle in Silicon The range of an alpha particle in silicon can be estimated within 15 per cent by the Bragg-Kleeman rule, RSi= 3.2 x 104 x x air a P a Si 210 air The range in air of a 5.3 Mev P a-particle is approximately R z 3.9 cm o aix For silicon, A = 28 and PSi = 2.33 gm/cm3. Thus Si (32 x 104)( 48)(5e9) (2533) 4 29.1 x 10- cm =29. 1 wm Now, for the detector under consideration, da 9 3m dd. 09 pm Including the junction depth (_, 1pm), the- depletion region extends approximately 94 pm into the detector. Thus, a po210 a particle will lose all its energy in the depletion region. The amount of ionization produced in the detector is proportional to the incident a-energy, as long as i <a da If i > da, ionization occurs outside of the depletion region. These e-h pairs may diffuse back into the depletion region-and be collected. However, most of 'these e-h pairs recombine before this occurs, 8.88

At what alpha energy would non-linearity be expected to occur? The answer to this question, of course, is that energy where the range in silicon exceeds the depletion depth. Using the Bragg-Kleeman relation Rair is determined by substituting RSfi 94 pm. Si a Psi ~Rai~l)r }.S. ' 1.23 cm. 03.2 x o0 4)Asil From Price (page 8) the energy of a~n a particle corresponding to a range in air of 13.2 cm is 11.4 Mev. Thus, with. a reverse bias of 100v, non-linearity would be expected to occur for a particles wThose incident energy is > 11.4 Mev. Of course, increasing the reverse bias would extend this energy limit. 8.;46 Energy Loss by Incident a-Particle A charged particle moving through material loses most of its energy by ionization. The energy loss per -unit path is given by the relativistic relation derived in an earlier part of this chapter, -dE 4je4z2NZ 2mov2 where z = +2 or +1 for alpha particle N = 0.50 x 1023 atom/cm3 for silicon Z 14 for silicon mo = 9.11 x L0328 gm =rest mrass of electron J 1 ioniza tioon potent. ial.4 eev v = velocity of charged partic le After the alpha particle slows down to a speed approximately equal to the speed of an orbital electron of a helium atom, the alpha picks up an orbital electron HeE+ + e H- e+. 8.47 Limiting Energy of Ionization (Ei) When the energy of an incident C is re'Iuced to some value Ei (called limiting energy of ionization), atomic displaceme-n:t7 replaces ionization as being the predominate method for losinig energy. The energy degradation formLla may be rewritten as =dE k dx J where~ k = 8r4z2Nz = 2mov2 8Q8

By setting the first derivative of the above with respect to a equal to zero, it can be shown that -dE/dx reaches a maximum at some alpha energy in ~ ] = 0 dx ca J 1 d.ln a/J + 1n a/J d(la O0 a dd d a 1 in (/J) + in a/J (-l/a2) 0 0 a dd lna /J a, dl aln /J a 1 d(0~J) - d, flc/J da a/J = e 2. 71 a 2m v2 2. 71 J The energy of the incoming a-particle is E.= 1/2 mnv2. Thus 2nMv2 = 2 2E = 2.71 J or m~ Em = (2, 71) J (- ) = energy at which maximum dE/dx occurs. 4 mO Also, as 2m0v + J the energy degradation formula approaches zero. -d.E k aI 0 as a J dx a a/J - 1 2mov -: Substituting in v2 2E/ma, l'"a 4 mo which is the energy at which ionization is no longer predominate. Most ionization occurs around the maximum point, which is near the end. of the a range 8 90

8.48 Time Required for a to Complete Ionization In order to get an idea of the length of time required for an a particle to ionize completely, the alpha range IRi is divided by the initial velocity of the a and then by the velocity where atomic displacement predominates. The time for ionization will lie in these limits. The initial velocity vo of a 5 Mev a particle is = 2E 2:x(5x 106)(1602 x 1012 gm cm2/sec) ma 66.69 x 10 g24m 1. 55 x 109 cm/sec = 1.55 x 107 -m/sec. ~ The energy at which atomic displacement becomes predominate is EF o 1 T-.,,5x.4 evx 1.602 x 10'12 vF =-2 1 -2 mXOc 9, 11 x 10'28 1.88 x 10+8 cm/sec 1.88 x 106 m/sec4 The minimum time is given by -6 29.1 x 10 m 18,8 x 10'13 19 0 Tmin 7M/ 1o9 x 10-1 ec Tmin 1. 55 x lO7msec Similarly, Tmax is -6 %Tax..29.l.x m 15. 5 x 10'12 - 1.6 x 10-11 sec rmax 1.88 x 106m/sec Thus, the time to complete ionization is less than 1,,6 x.10'11 sec and more than 1.9 x 10-12 sec. 8.49 Estimation of Charge Collection Time: and Charge Collected Neglecting diffusion currents, and assuming a constant average electric field in the depletion region, the total current density is given by J e in n g + e p Up 8,91

Let V volume of depletion region, A - cross sectional area and d = deplation.depth. The total current is given by N P I = AJ = e - In A + p A V V. p The charge collected in time At is N P mQ = At = ett (d -nt + dE ) Now, vn and and vp = p thus, Ax - Xn At v= nAt I1n~At n At At and p LX 1At v At 11 pLThe total charge collected is AQ Ne (xn + Axp) d If all electrons and holes reach the electrodes, Axn xn and.p:Z Xp and (xn + Xp) AQ~ =Ne =Ne d since xn + xp = d, where xn is the distance from the point of e-h formation to the p electrode and xp is the distance to the n electrode. If the e-h pairs d.o not reach the plate, the following inequality holds Q = Ne(x_+,) <Ne d The maximnum time required for an electron to induce a11 its charge is determined in the following manner. Assume the e-h pair is formed just ink side the p electrode, The electron will migrate to the n electrode under the effect of the electric field. The time for an electron to do this gives an indication of how fast the charge is collected (this is a lower limit). Let the electric field be ta = 105 v/cm, and the electron mobility be An = 1300 cm2/V sec. Then the average velocity of the electron is vn -n tav 1 3 x 108 cm/sec.o The time to travel the width of the depletion region d z 93 x 104 cm is T. 93 x 104 cm l2 x 1012 11l 1,3 x 108 cm/sec 8~92

8. 50 Estimation of Voltage Output of the Detector Assume the following~ 1. Detector biased.100 v negative. 2, Barrier layer capacitance Cd 2 65 4yf/mm2 x 2 mm2 =5 3 ~f 210 3. Incident Po ac particle (Eo = 5,3 Mev) loses all energy in depletion regiono 4. 3, 6 +.3 ev/e-h pair = Energy of Ionization, 5, Input capacitance of electronic system = 60 4If = Ca The output pulse is given by V - out (Cd + Ca) coulomb /AQ - Ne 5 x ev (' 602 x 10- 2.36 x 106 3 3.6 ev/e-h pr. C Cd + Ca = (.27 + 6o) 4 =f 60.3 P rf 2,36 x lO0'le couo mb -1 Vout 6o - m o 039 x 10 volt = 3.9 mv. 60~ 3,url Thus, the expected pulse voltage would be less t;hlan 4 mvy 8.93

BIBLIOGRAPHY L McKay, K. G., "Phys. Rev.," 84, 829 (1951). 2. Spenke, E., Electronic Semiconductors, McGraw-Hill, Inc. New York, 1958. 3. Smith, R. A., Semiconductors, Cambridge University Press, London, 1959. 4. Mayer, J. W. and B. R. Gossick, "Rev. Sci. Instru.," 27, 407 (1956). 5. Walter, F. J. J. W. Doobs, L. L. Roberts, and H. W. Wright, ORNL 58-11-99 (1958). 6. McKenzie, J. M. and D. A. Bromley, "Phys. Rev. Letters,t 2, 303 (1959). 7. Friedland, S. S. and J. W. Mayer, "Nucleonics," 18, No. 2, 54 (1960). 8 94

8.51 Experiment on Alpha Particle Detection Using the P-N Junction Purpose To obtain differential pulse height spectrums for Po210 alphas using the P-N junction detector under several different external biasing conditionso To obtain from the above, energy resolution as a function of bias, pulse height as a function of bias and detector voltage output. To determine the gain characteristics of the transistorized preamplifier and its input noise signal. Incident alpha particles lose their energy in the detector predominantly by ionization producing electron hole pairs. Stror electron and hole concentration gradients exist in the junction region which result in a changing space charge density and its attendant electric field. Due to this strong electric field ( - 105 v/cm) holes migrate to the p-type material and the electrons to the n-type material resulting in a collection of charge which along with the junction barrier capacitance and input capacitance to the amplifying system produces a voltage pulse on the order of millivoltso The charge collection time is known to be less than 3~5 x 10-9 seco List of Equipment 1 m curie Po210 alpha source (T1/2 = 138 d, E = 5~3 Mev) Hughes P-N junction detector Transistorized pre-amplifier (Rise time = 0o2s, Gain = 45) Linear Amplifier (TR = 0~2 t sec) Single channel differential analyzer High voltage power supply Scaler counter Pulse height generator (0o2 t sec T or less, millivolt range) Cathode ray oscilloscope (at least 0~2 t sec rise time) It should be noted that the display of the signal voltage pulse is limited by the electronics. It would be desirable to. have electronics which approach 10-9 sec rise time or lesso Procedure lo Using the p210 a source (1 mc, 5~38 mev) and the detectorpreamp-amp-difo anlo system, obtain differential spectrums 8. 95

for several detector reverse biases (0 to 180 v Dc) i.e., for a given window width, obtain count rate for various Edial settings about the -peak count rate. Plot relative pulse height vs. count rate for each of the different biases Rel. Rs 100o x full width at 1/2 max pulse height Count % Res Rate Energy under maximum P.H. 1/2 = -— E x 100% Ea Ea dEdial Figure 8.33 Relative Count Rate vs. E Dial. Compute the energy resolution for each of the different biases. Plot resolution vs. detector bias. %Res. - Detector Bias Figure 8.34 Percent Resolution vs, Detector Bias. From the E.'s given from above curves, plot detector bias vs. pulse height at E. pulse height / at Ed V = Q/Ca + CDET Detector Bias Figure 8.35 Pulse Height at E. vs. Detector Bias. 2. Determine the input noise to the detector-preamp-amplifier system utilizing the CRO Voltage Out of Amplifier System Input Noise (Gain of Preamp)(Gain of Amp) 8.96

3, Determine the voltage output from the detector and thus, the input signal to noise ratio. Signal Pulse Voltage Out of Amp Detector Output Voltage (Gain of Preamp) (Gain of Amp) Detector Output Voltage Signal to Noise Ratlio System Input Noise 8.52 Questions The following questions are representative of those used in exams covering the topics considered in this chapter. Your background should be broadened by outside reading until this type of question, which stresses qualitative understanding of the equipment and the processes involved, is no challenge, 8o53 Ionization Chambers 1. Given mean level ionization chamber which has been calibrated using a radium source of 5 curies: (a). Discuss the problems involved, qualitatively, in determining the strength of any radioactive material in curies using this calibrated instrument. (b). Using formulae if necessary, describe carefully the specific method you would use to convert the measurement of the radioactive material using the calibrated instrument to the exact strength in curies. 2. Consid.r ibe following statement in connection with ionization chambers. "Very large chambers are not practical, and therefore an air equivalent wall is used to increase the sensitive volume of the chamber and to increase its efficiency. The wall is made of materials with about the same average atomic mass as that of air. The thickness of the wall must be less than the range of the ions which are formed inside the wall so that these ions can get to the collecting electrode and produce discharges. Discuss any inaccuracies, 3 An ionization chamber and a G-M counter are operating properly with positive high voltages or their center wires and their shells grounded, Would you expect them to operate properly if one would change the 8 97

sign of the voltages impressed upon the center wires? Please give reasons for your answer. 4. You are given the task of designing arn air-equivalent ionization chamber. In selecting the wall thickness, what considerations will govern your decision? 8.54 GM-Proportional Counters 1. A radiation survey is to be made of the gamma field in a radiation laboratory. Using a Geiger-Muller type survey instrument, Mr. X performs this task by reading the mr/hr figures indicated by the meter of the instrument. Under what conditions, if any, are the results of this survey meaningful? Please state fully -the reasons for your answer. 2. What determines the resolving time of the GM counting system? 3. What determines the resolving time of the proportional counting system? 4. (a) Discuss qualitatively the basic concepts required to develop the theoretical form of the voltage pulse from a GM tube. (b) Discuss the relative magnitudes of the migration times of the positive and negatively charged particles, t and t_, respectively, in a pulse ionization chambers. Be specific in considering (1) air equivalent type chamber, (2) proportional chambers and (3) GM tubes. (c) Draw a representative pulse for a GM tube when the time constant, RC, is (1) RC < t_, (2) t_ < RC << t, and (3) t < RC. 5. For a GM tube: (a) Draw the voltage pulse forms as seen with a CRO to determine dead time, resolving time and recovery time. Indicate on your figure how these times would be measured. (b) Draw the voltage on the collecting electrode as a function of time corresponding to the figure in part a. Explain. (c) Draw the resolving time of the GM as a function of anode voltage. Explain the behavior you have shown physically. 8.,98

6. (a) In the gas flow proportional counter, what specific characteristics of the gas are desirable? (b) Is a quenching gas necessary, as in the GM tube? Explain. (c) What effect does the gas pressure have on the overall multiplication? What considerations would effect the operating gas pressure of a chamber which you had the responsibility to design? 8&55 Statistics 1. Given: A radioactive source whose disintegrations satisfy a Poisson population description, and a GM counting setup. (a) Considering the low efficiency of the GM system for detecting gammas, justify the use of a Poisson distribution when analyzing the sample data. (b) Discuss the effect` of the source gamma energy spectrum on the analysis of this sample data (c) Discuss the effect of geometry and source activity on this analysis~ 2, Prove or disprove that the sample mean is known with less uncertainty when the data is taken in 10 separate, equal time intervals, T, rather than in one interval, 10To 3. Prove or disprove that the variance of the data is identical for both methods of data taking in part (d), when the total. counts recorded are identical. 40 (a) What is the difference between sample and population. data? (b) Are statistical quantities such as standard deviation. derived from sample data information? 5. Please state the physical requirements, and only those physical. requirements, which have to be satisfied in order that one may apply the following statistical. laws to radioactive decay. (a) Poisson s distribution (b) Gaussian. distribution 8,99

Do not use mathematical symbols and inequalities, and do not make your statements redundant. 6. You take two successive one minute counts of a radioactive material. The first yields 10,000 count, the second 9,500; please comment very briefly on these results. 7. Student A counts a source for 10 minutes. Student B counts a similar source for 10 one minute intervals. Student A remarks to B: "Since both of us obtained about the same total number of counts and therefore our results have about the same standard deviation, you certainly wasted a lot of time taking your data in ten pieces, and by resetting and restarting your scaler." Please discuss this statement. 8.56 Scintillation Spectrometry 1. You are given the task of determining the absolute activity of a small amount of radioactive material of the order of one microcurie, which emits only monoenergetic gamma radiation of about one Mev energy. No calibrated standard of this material is available. (a) By means of a detailed block diagram, show the pieces of equipment you need. Your diagram should show a separate block for each function to be performed, that is, you might have to show several blocks for pieces of equipment built on the same chassis. Carefully label all blocks and interconnections. Describe fully the function of all blocks and interconnections. (b) Describe fully the measurements you would make with -this equipment on -the radioactive material. (c) Describe fully how you would use the results of your measurements to compute the activity of the material. If information in addition to your experimental results is needed, please state clearly what this information is. 2. Given: a monoenergetic y source and a shielded well type NaI(Tl) crystal and a single channel analyzer setup operated in differential. (a) Explain the reason for the presence of a backscatter peak and its corresponding energy. (b) What is the effect of the size of the crystal on the relative height and width of this peak? Explain. 8,100

(c) Explain the decrease in count rate to the right of the Compton edge. (d) Physically explain the observed photopeak resolution of 10o%. (e) How is the photopeak resolution affected by the size of the crystal? Explain. 3. For a source with an equal number of gammas per unit energy and a continuous energy distribution up to a cut-off of 1.02 Mev, draw a representative curve of' count rate vs. Edial (where Edial should be plotted in equivalent gamma energy) for a NaI(Tl) well crystal on INTEGRAL. Explain-the behavior. 4. Repeat part c for DIFFERENTIAL. (Cont'd,) 5. What differences should one expect to see in the differential spectrum for a given gamma source between a well and a crystal. scintillation counter. Explain. fully. 6. What is the Compton edge? What is its significance in analyzing a differential spectrum from a mixed gamma source? 7. Consider a gamma of energy E, E > 102 Mev, impinging on a NaI(Tl) crystal In pair production, the kinetic energy of each of the two betas formed is (E - 1.02)/2. When the positron recombines with an electron (assumed at rest) the kinetic energy of either of the resulting gammas (moving in opposite directions) is just (E - 1.02)/2 + 0.51 = E/2. Assume the resulting gammas can lose all their energy in the crystal, or one or both can escape entirely, the pair production photopeaks should occur at E - E/2 and E - 2(E/2) = 0 rather than at E - 0.51 and E - 1.02 as commonly claimed. Please make a critical analysis of the above remarks, being sure to explain any points of disagreement., if any. 8.57 Neutron -Detection (a) What is meant by the optimal or correct settings for the anode voltage and the discriminator and the amplifier gain? (b) Draw on the same graph representative curves of count rate vs. discriminator setting for- lo the optimal anode voltage, 2. a higher voltage, and 3~ a lower voltage. Explain this behavior. (c) Draw representative curves of count rate vs. anode voltage with 3 discriminator settings as parameter on the same graph. Use 8 101

the correct discriminator setting for the optimal anode voltage and a higher and a lower discriminator setting. Explain this behavior. (d) Can the CR0 be used to find the correct discriminator setting if the optimal anode voltage is known? Explain. (e) Can the CR0O be used to find the optimal anode voltage if the proper discriminator setting is known? Explain. (f) Can the CR0 be used to find the proper settings for both discriminator and anode voltage given a BF3 tube about which no characteristics are known? Explain. (g) Define the efficiency of a BF tube for neutron detection. What is the order of magnitude of this efficiency? Why? (h) Why are the neutron induced pulses from the BF3 tube not of uniform height? (i) In the BF3 experiment, the importance of keeping the field at a low (e.g. less than 1000r/hr) value was emphasized. How could you design a gaseous BF3 tube which reduces the effect of gammas on tube operating characteristics while maintaining the same neutron sensitivity? Explain. 8.58 Foils 1. In the foil activation experiment, we speak of a saturation activity. Explain this term physically. 2. Is it necessary to know the saturated activity to calculate the relative flux? Explain. 3. How could you use the foils to find absolute flux levels? 4. What characteristics does an ideal foil have? 5. What considerations determine the counting equipment you would use to measure the radiation given off by the foils? 8.59 Solid State Detector 1. What effect does an increase in reverse bias have on the space charg 8102

density, the electric field, depletion depth, barrier capacitance and output pulse from an a particle? (Be specific, give equations and sketches to explain)o 2. Suppose the reverse bias is removed from the detector while the detector is still. connected to the electronics. Will it still be possible to detect incident alpha particles? Explain. 3. What effect does extending the high frequency range seen by the electronics have on the observed alpha voltage pulse? 4. Why does the pulse height vs. alpha particle energy curve become nonlinear? 5. If the junction were formed 12 microns beneath the surface, what effect would this have on the pulse height vs. alpha particle energy curve? 6-. Why is the solid state detector more effective than a crystal detector with an externally applied electric field? 7. Is there any advantage to having a detector with a resistivity of 100,000 ohm-cm rather than the 300 ohm-cm detector used to take the data in this laboratory? 8. In the solid state detector experiment, the pulse at the amplifier output was observed to reach a saturation value at a reverse bias of about 30 volts. Why? 9. Why is the observed pulse rise time about one microsecond? 10. Why is the detector light sensitive but not effective in measuring our beta sources (0.1 Lc). 8.60 General 1. Given a current ionization chamber calibrated against a known radium source in mr/hr. What effects must be considered when using this chamber to measure the STRENGTH of an unknown gamma source. Be specific, using equations. 2. Given a calibrated Geiger-Mueller chamber, instead of a current ionization chamber, reanswer (1)o 8Blo3

3. Given: a source that emits a, A, and y (a) What equipment would you use to count these particles, under the restriction that the counts due to each type of particle be resolveable? Draw a block diagram(s) of the proposed system(s). (b) How could absolute counting be done using the equipment of part a? 4. Comment on the relative merits of an end window GM counter and a well type NaI scintillation counter for determining the activities of neutron activated materials, such as indium-116 and gold-198, which emit both betas and gammas. Consider such points as counter efficiency, background, and effect of foil thickness. 5. Please describe fully the physical phenomena which account for the rise in the output current, or count rate, with increased voltage in the plateau regions of the following instruments: a) Ionization chamber b) Proportional counter c) Geiger counter d) Scintillation counter (on integral count) 6. (a) Explain the difference between the "total absorption coefficient" and the "true energy absorption coefficient" for gamma radiation. (b) Which of the above mentioned coefficients would you expect to obtain by measuring the attenuation of a narrow beam of gammas? Please explain why. (c) Which one of the above mentioned coefficients would you use in calculating the radiation dose rate in roentgens per unit time at a certain distance from a given radiation source? Please explain why. 7. Develop and explain clearly the theory of the two source method of resolving time measurement. State the approximations involved and state approximately what the counting rates should be for best results. Do not restrict your discussion to one type of detector. 8. Recommend a type of instrument to de each of the following jobs. Explain in detail your reasons. 8 104

(a) Coniduct tracer analysis using a gamma emitting isotope. (b) Monitor the radiation level in the open air after an atomic attack. (c) Measure exposure dose at a few feet from a 20 mc unknown gamma source. (d) Measure exposure dose at a few feet from a 20 mc Co-60 gamma source. (e) Monitor continuously the effluent gases from ventilation waste stacks from a laboratory using many different kinds of radioactive materials. 8.61 Calculational Problems 1. Given a known point source of C curies: (a) SYMBOLICALLY, develop the formalism for computing the whole body dose in REM which you would receive standing ten feet from the exposed source. Define all SYMBOLS CLEARLY, giving UNITS. State ASSUMPTIONS. (b) At a distance for which your survey meter is on scale, SYMBOLICALLY relate the instrument reading in mr/hr (calibrated with a RADIUM source) to the actual air dose rate at the location of the meter caused by the hot source to which you are exposed. (c) At this distance, SYMBOLICALLY, develop a GENERAL relation expressing the dose rate which your body receives on the basis of the measured meter reading. (d) If the instrument in part b had been a Geiger Mueller rather than a mean level type, please reanswer part b. (e) For the GM meter, reanswer part c. 2. Develop a GENERAL expression relating the strength of a point gamma source in curies to the dose rate, r/hr, at a distance d cm. in air. Define all terms used. Do not restrict to one gamma per disintegration, nor to one gamma energy. (a) Write an integral expression for the rate with which neutrons interact with a BF3 gas tube for a given geometry. Define all terms used. State any important assumptions clearly. 3 i03

UNIVERSITY OF MICHIGAN 3 9015 02539 7285 3. One of the men under your supervision in a radiation laboratory accidentally swallows some radioactive material. Upon investigation you find that this particular isotope emits gamma radiation only, becomes evenly distributed in the.body and is eliminated from the body in an approximately linear manner in about 100 hours. The material undergoes 3.72 x 1011 disintegrations per hour, emitting a 0.625 Mev (=10-6 erg) gamma per disintegration, with a half life of 2.2 years. About half of the gamma energy is absorbed in the body. (a) Compute the total radiation dose in REM units that this man will receive from this material. (b) Decide on the period of time this man should spend away from his regular job where he receives 2.5 millirems per hour on the average. (c) One of the other men under your supervision remarks: "It would have been much better if the material this fellow swallowed had been an alpha emitter instead of a gamma emitter. The alphas, due to their small penetrating power, would cause much less damage than the gammas." Please comment on this remark. 4. Derive an expression for the dose rate,.D, in mr/hr, at a distance d (cm) from a point source of strength S (curies). Define all conversion factors required SYMBOLICALLY, being SPECIFIC on the UNITS of all symbols used. 5. Radium gives off a variety of gammas per alpha disintegration. Qualitatively, how is this effect exhibited in the expresson derived in 4-, given S curies of radium? 6. Show SYMBOLICALLY `he procedure which you we d follow to arrive at the present strength of a Co60 source, given the dose rate D in mr/hr at a distance, X cm at a prior timne T, such as March 20, 1952. Define any additional symbols raqaired in your analysis as needed. 8.106