UNIVERSITY OF MICHIGAN COLLEGE OF ENGINEERING INDUSTRY PROGRAM NUCLEAR ENGINEERING ENGINEERING APPLICATIONS OF NUCLEAR ENERGY D. L. KATZ H. A. OHLGREN KENT DEDRICK J. G. LEWIS MARX WEECH April, 1955 IP-115

ACOKNOWLEDGMENT The authors are grateful to the Atomic Energy Commission for supplying much of the information used in preparing this publication, and they also wish to acknowledge the support of the College of Engineering Industry Program and the valuable assistance given by many members of the faculty and staff at the University of Michigan Engineering Research Institute under the direction of R. G. Folsom.

TABLE OF CONTENTS Page PART I Nuclear Physics 1. The Atomic Nucleus 1 A. Introduction 1 B. Chart of the Nuclides 2 C. The Binding Energy Curve 4 2. Radioactivity 7 A. Alpha Particle Decay 7 B. Beta Decay 8 C. Gamma Ray Emission 11 D. Natural and Artificial Radioactive Nuclei 11 3. Nuclear Reactions 17 A. General 17 B. The Compound Nucleus Theory of Nuclear Reactions 18 C. Nuclear Fission 19 D. Fusion and Thermonuclear Reactions 26 E. Nuclear Cross Sections 27 PART II Nuclear Chain Reaction - Reactor Materials 1. Chain Reactions and Nuclear Reactors - General Discussion 33 A. Nuclear Chain Reaction; the Critical Fuel Mass 33 B. Reactor Classification Based on Neutron Energies 37 C. Reactor Classification Based on Fuel Distribution 37 D. Reflectors and Blankets 38 2. Nuclear Fuels 38 A. Natural Uranium 38 B. Thorium 45 C. The Transuranium Elements - Plutonium 46 D. Fuel Enrichment 49 E. Fuel Breeding 51

Page 3. Reactor Control and Reactor Materials 54 A. Reactor Control 54 B. Moderators 58 C. Coolants 63 D. Structural Materials 65 E. Radiation Damage 69 F. Shielding 70 Part III Nuclear Engineering 1. Reactor Types and Fuels 85 A. Thermal Reactors 85 B. Intermediate Reactors 86 C. Fast Reactors 86 D. Reactor Fuel Elements and Fuels for Homogeneous Reactors 88 2. Power from Nuclear Reactors: The Nuclear Manufacturing Plant 89 A. Introduction 89 B. Power from Nuclear Reactors 90 C. The Nuclear Manufacturing Plant 93 3. Fuel Separation 103 A. Introduction 103 B. Aqueous Methods 104 C. Fluoride Volatility Methods 107 D. Pyrometallurgical Processing 107 4. Use and Storage of Fission Products 109 A. Introduction 109 B. Storage of Fission Products 111 C. Use of Fission Products 112 5. Conversion Factors 114

LIST OF TABLES Table Page 1 The Uranium Series 13 2 Common Radioisotopes 14 3 Delayed Neutrons 22 4 Thermal Neutron Absorption Cross Sections 30 5 Sources of Uranium 40 6 Thermal Neutron Fuel Metals Cross Sections 44 7 Uranium Family 47 8 Thorium Family 47 9 Plutonium Family 47 10 Some Possible Types of Power Reactors 53 11 Properties of Moderators 60 12 Deuterium Separation Processes 62 13 Properties of Coolants 66 14 Properties of Structural Materials 68 15 Gamma Ray Absorption Coefficients in cm"for Iron, Lead, and Concrete 74 16 Reactor Gamma Rays 74 i*7 Neutron Shielding. Data 77 18 Heavy-Aggregate Concretes and Other Shielding Materials 78 19 Radiation Tolerance Levels 83

LIST OF FIGURES Figure Page 1 Chart of the Nuclides 3 2 Binding Energy per Nucleon for Stable Nuclei 5 3 A Typical Beta Decay Spectrum 10 4 The Fission of a Typical U-235 Nucleus 20 5 The Fission Process 21 6 Spectrum of Neutrons Produced by the Slow Fission of U-235 and Pu-239 24 7 Fission Product Mass Spectrum 25 8 The Total Neutron Cross Sections for Some Reactor Materials 52 9 Determination of the Critical Mass by the Critical Assembly Experiment 36 10 Basic Reactor Schematic 39 11 Specific Heat of Uranium at Elevated Temperatures 43 12 Nuclear Reactions Leading to the Build-up of the Transuranium Elements 48 13 Gaseous Diffusion Cascade 50 14 Electromagnetic Isotope Separation 50 15 Build-up of Fissile Fuel and Fission Products in Thorium and Uranium in a Nuclear Reactor 55 16 Elastic Collision of a Neutron with a Moderator Nucleus 59 17 Comparison of Fluids for Heat Transfer 67 18 The Maximum Range of Beta Particles as a Function of Energy 72 19 Nomograph for Determination of Reactor Fuel Requirements 76 20 Nomograph for Calculation of Energy Available from Radioactive Nuclides 79 2 21 Gamma Ray Flux in Photons per cm -sec. which Produce One Roentgen Per Hour in Air 80

Figure Page 22 Approximate Permissible Neutron Flux Based on an Exposure Time of 40 Hourse Per Week 84 23a Heterogeneous Thermal Fissioning Reactor 87 25b Homogeneous Thermal Fissioning Reactor 87 23c Heterogeneous Fast or Intermediate Energy Fissioning Reactor 87 23d Heterogeneous Fast Breeder or Converter Reactor 87 24 Fuel Reserves and Estimated Future Energy Comsumption 91 25 Enriched Uranium Reactor System with Enrichment Plant 95 26 Plutonium Re-cycle Reactor System Using Natural Uranium Fuel 97 27 Fast Plutonium - U238 Breeder Reactor System 99 28 Thermal U23 - Thorium Breeder Reactor System 101 29 Solvent Extraction Process for Irradiated Uranium 106 30 Block Flow Diagram of a Fluoride Volatility Process 108 31 Block Flow Diagram of a Pyrometallurgical Process 110 32 Possible Uses for Fission Products 113

PART I - NUCLEAR PHYSICS 1. THE ATOMIC NUCLEUS A. Introduction The atom is often described as the smallest amount of material that has a definite chemical identity. An atom consists of a positively charged central core, the nucleus, and a surrounding cloud of orbital electrons. The number of orbital electrons is equal to the number of positive charges on the nucleus so that an atom is electrically neutral. The electrons may occupy a series of different orbits of gradually in. creasing average radius but once an orbit contains two electrons, no more can be accepted. The chemical properties of an element are determined by the electrons in the outermost orbits. Most all atoms are about the same size with diameters slightly greater than about 108 cm. The atomic nucleus is small and very dense. Nuclear radii are all roughly 1013 cm, the heavy nuclei being somewhat larger. About 0.02% of the mass of an atom is due to the mass of the orbital electrons, the remainder being concentrated in the nucleus. The nucleus is believed to consist of Z protons each with unit positive charge, and N neutrons which are- neutral particles. The number Z is the atomic number. The chemical properties of an element are determined by Z and so the chemical periodic table is built up with Z as an index numbero The number of neutrons -N does not influence chemical behavior, but does influence nuclear properties considerably. Different nuclei or nuclides having the same Z but different N are A nuclide is a particular nuclear species with given values of Z and mass number A = N + Zo Nuclei refer to more than one nucleus and may include a number of nuclides. Where no confusion is likely, nuclei and nuclides are used synonomously in the literature. 1

called isotopes. The mass number of a nucleus A, is the total number of neutrons and protons. Nuclides having the same A but different Z are called isobars. Neutrons and protons have very nearly equal masses of about 1.67 x -24 10 grams, the neutron being heavier than the proton by about 0.1 percent. The mass of a nucleus is slightly less than Z x (mass of a proton) + N x (mass of a neutron), The difference or mass defect, being about 0.8 percent of the total mass for most nuclideso A more detailed description of the mass defect is presented in Part C of this section. B. Chart of the Nuclides The chart of the nuclides is a plot of Z vs N (see Figure 1). This chart shows both stable and radioactive nuclides and gives information about the modes of decay. Excellent wall size copies of the chart may be obtained from the Knolls Atomic Power Laboratory, Schenectady, New York. The stable nuclides are closely clustered about a curved line running across the chart. The radioactive nuclides generally fall at some distance on either side of this line, and the schemes by means of which they decay into stable nuclides may be traced on the chart. A shorthand notation has been developed to specify particular nuclides. The chemical symbol for the element, the atomic number Z, and the mass number A are written in the followg coing combination Z(chem. symbol)A p206 For example, 82P means the lead nucleus with a mass number of 206. All lead isotopes have 82 protons. The number of neutrons in 82Pb206 is A-Z = 124. Often the value of Z is omitted in this notation as the chemi20o6 cal symbol fixes Z, and we see Pb, Pb-206, and lead-206 in the literature. 2

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C, The Binding Energy Curve When two or more nucleons (nucleon is a generic term which includes both neutrons and protons) are in close proximity, there is a not too well understood short range force between them which will cause them to form a cluster. It is an energetically favorable situation to have nucleons so clustered as a nucleus. The amount of energy needed to break a nucleus into neutrons and protons infinitely separated from one another is called the binding energy for that nucleus. The binding energy HE can be measured by noting that by the Einstein relation between mass and energy, we expect a nucleus to weigh less than its constituent nucleons by an amount AM = ZE/C2. Measurement of nuclear masses by means of a mass spectograph* provides a measurement of the mass defect and consequently, the binding energy. Binding energies are usually stated in Mev, (million electron volts) although other units of energy may be used. The conversion tables on page 113 permit conversion to B.T.U,, kilowatt-hours, etc. The binding energy LE is very closely proportional to the mass number A, thus we find there is a nearly constant value of the binding energy per nucleon ( E/A). The deviations from proportionality show up best in a plot of ( AE/A) vs. A (see Figure 2). The mass spectrograph is an instrument which measures the mass of a charged particle, in this case an ion of the element whose mass is to be determined. The ion is given a definite velocity by accelerating it with a known electric potential and then allowed to pass through a region where an intense static magnetic field has been set up. The ion follows a circular path in the magnetic field. The radius of the path is different for ions of different masses and so measurement of the path radius provides a measurement of the mass of the ion. **.The:ergy unit Mev is the kinetic energy which an electron has after falling through an electrical potential of one million volts. One Mev may be seen to be equivalent to 1.6 x 10-12 ergs. The Mev unit of energy is not restricted in use to describe electrons.

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This curve is important to nuclear power engineering in that it shows how much energy may be obtained from a heavy nucleus like uranium by its being broken into two or more fragments by the fission process. The energy available from each fission is given by: (sum of binding energies of fission fragments) - (binding energy of the initial uranium nucleus). For uranium, this amounts to about 200 Mev for each fission. A little consideration shows that the most stable nuclei are those with the maximum value of (AE/A) and hence in the middle of the curve of Figure 2. It is energetically favorable for some light weight nuclei to combine to form a medium weight nucleus by the fusion process and it is also energetically favorable for the very heavy nuclei to break up into medium weight fragments. The alpha particle or 2He nucleus is a light weight nucleus of high stability as may be seen from Figure 2. This is important in the theory of alpha particle emission by heavy nuclei and will be discussed in the next Section. The high binding energy of the alpha particle is also the basis of the explanation of the energy output of the sun where it is believed that the initial reactants are hydrogen and light nuclei and the products are helium and other light nuclei. The overall process being the conversion of hydrogen to helium with the release of about 28 Mew for each helium nucleus produced or 0.75 megawatt-hours for each mole of helium produced. Published values of measured nuclear masses are expressed in atomic mass units, (amu). This unit has been chosen such that oxygen - 16 has a mass on this scale of 16.000. The atomic mass unit may be shown to be 1.660 x 10-24 grams. The chemical mass scale is very nearly the same as this but uses naturally occurring oxygen as a basis instead of oxygen 16. 6

The conversion tables give the conversion factor between these mass scales. 2.o RADIOACTIVITY A radioactive nucleus is one that is energetically unstable and if given suIfficient time will decay to a more stable state by the emission of either a particle, a gamma ray, or both. The probability of the decay of a particular radioactive nucleus is independent of the histories of all other radioactive nuclei in a sample of material. This leads to the result that the rate of decay of a sample of radioactive nuclei is given by dX -=-; or X = Xoe. (1) Where X is the number of radioactive nuclei which have not decayed at time t, Xo is the value of X at t = 0, and A is a number characteristic of the radioactive nucleus in question and is related to the half life. The half life T1/2 is the time required for half of the nuclei to decay. From equation 1, we see that T12 0.693. The mean life 3of a radioactive Fromn equation 1, we see that Ti/2 a radioat. nucleus is the average time a nucleus can be expected to spend before decaying. This can be seen to be given by'= i. In terms of, we have T1/2 = (0o693)'. The times for half lives.and mean lives may be measured in any convenient unit. A. Alpha:.Particlel Decay Many heavy nuclei are unstable against decay by alpha: particle emission (see Figure 1). The half lives of naturally occurring alpha particle emitters are generally inthe neighborood of theousands to millions of years thus accounting for their natural existence in the earth's crust. It is possible by means of the binding energy curve of Figure 2 to determine if a particular nucleus is stable against alpha decay, the condition being that the sum of the binding energies for the daughter nucleus 7

and an alpha particle be greater than the binding energy of the parent nucleus. Any excess of energy then manifests itself as the kinetic energy of the alpha particle and perhaps also some energy will go into the production of one or more gamma rays. The alpha particles have well defined energies which generally fall in the range 3 to 10 Mev. Alpha particles have well defined short ranges and by themselves present no hazard to workers unless the source is in contact with the skin or the radioactive material is taken internally. The gamma rays associated with the alpha particle decay of some nuclides may be dangerous at a distance however. B. Beta Decay The beta rays discovered and investigated by Becquerel, the Curies, and others near the year 1900, have been identified as being high energy electrons. Anderson in 1932, discovered the positron or positively charged electron in cosmic radiation. Shortly later, positrons were observed to be emitted in the decay of certain artificially produced radioactive nuclei. It has been observed that the details of the radioactive electron and positron decay processes are very similar. Further, it is felt that the K capture and L capture processes in which an atomic electron in a K or L orbital is absorbed by a radioactive nucleus are closely related to the electron and positron emission processes mentioned above. Therefore, we often see these three different kinds of radioactive decay referred to as beta processes. In beta processes, the nucleus changes its atomic number Z by one unit, but the mass number A does not change, or we may say a neutron has changed into a proton or vice versa. In the case of electron emission Z increases by unity, and in positron emission and K and L electron capture, Z decreases by unity. We often see electron and positron decays represented 8

in the following ways: electron 1H3 - 2He3 or 1H3 e 2He3 el1 12y 2 12y emission H3 -2He3 + e- (12y), etc. N13 + C13 o N13 e+ C13 positron 7 10 min 6 7 10 min emission N13 - C3 + e+ (10 min), etc. 7 6 Simple criteria may be readily developed to determine if a nucleus is stable against beta decay. These, as in the case of alpha particle decay, depend on the binding energies of the parent and daughter nuclei, and on the amount of energy necessary to create the beta particle. (0.511 Mev) The energy change Eo in the process has a definite value, yet a spectrum of electron or positron energies is observed from zero up to Eo. Figure 3 shows such a spectrum. Fermi and Pauli postulate the existence of a neutral particle, the neutrino, which is emitted in coincidence with the electron or positron and carries away an amount of energy such that the Sa= of the beta particle energy and the neutrino energy is equal to E. The tabulated beta decay energies are the values of E0. Neutrinos interact so slightly with matter that they usually escape the region in which they were produced. Considerable energy is lost from nuclear reactors in this way and amounts to about 5 percent of the fission energy of uranium or plutonium. The half lives of beta radioactive nuclides vary from fractions of a second to many years. Beta particles may be stopped by thin metal sheets. Their ranges in material are not as well defined as in the case of alpha particles. Figure 18 in Part II, Section 3-F shows the thickness of various materials necessary to stop beta particles of various energies. 9

w.E _ w Q..S~~YS Is jh)~ Wz JaJ d | / fD7 m w A9U 3N3 INn H3d S31011tlVd V138 40 H:38w'nN

C. Gamma Ray Emission Gamma-rays are electromagnetic radiation like X-rays but generally have much shorter wave lengths. Gamma rays produced in nuclei are monochromatic and have quantum energies between several kilovolts up to nearly 18 Mev. A nucleus radiates a gamma ray in transitions between excited states just as occurs in the emission of visible light by atoms. It is evident that neither Z nor A changes in the radiating nucleus. Gamma rays do not have well defined ranges but are absorbed exponentially (see the discussion under shielding). The rate of absorption is greatest for heavy metals. Alpha and beta particle emitters usually emit gamma rays as well. After the emission of the alpha or beta particle, the daughter nucleus is left in an excited state which decays almost instantaneously (within about -12 10- seconds) with the radiation of a gamma ray. The products of nuclear reactions are often left in excited states and also radiate gamma rays. Some gamma radioactive nuclei have lifetimes much longer than the usual value of about 10-12 seconds. Nuclides in these long lived excited states are called nuclear isomers and decay by isomeric transitions (often abbreviated I.T.) to the ground state. D. Natural and Artificial Radioactive Nuclei The vast majority of naturally occurring radioactive elements are heavy metals at the far end of the periodic table (see Figure 1). There are three distinct naturally occurring decay chains of heavy metal radioactive nuclei. These are the Uranium Series which starts with 9238 and after many alpha and beta decays, results in the formation of stable 82Pb206, the Thorium Series which has 90Th232 as a precursor and eventually yields stable 82Pb208, and finally the Actinium Series with 9235 as a precursor and stable 82Pb207 11

as an end product, the element actnium being an intermediate product. The details of the Uranium Series decay chain are given below in Table 2 as an example.* There is another series of heavy metal radioactive elements similar to the three above series which has been observed in the decay of artificially produced radio-nuclides but has not yet been found to occur in nature. This is the Neptunium Series which starts with 94Pu241 and ends with stable 83Bi209 A selected list of both naturally occurring and artificially produced radioactive nuclides is given below in Table 2.** More detailed compilations of alpha and beta decays are given in: "Table of Total Beta-Disintegration Energies" by R. W. King, and"Table of the Alpha-Disintegration Energies" by F. Asaro and I. Perlman in Reviews of Modern Physics, October 1954. Reprints of these articles may be obtained from the Publications Office, National Research Council, 2101 Constitution Avenue, Washington 25, D. C. * This decay chain as well as the others are described in detail in Sourcebook on Atomic Energy, by Samuel Glasstone, D. Van Nostrand Company, Inc., New York, 1950, p. 125. ** Taken from Introduction to Nuclear Engineering, by R. Stephenson, McGraw Hill Book Company, Inc., New York, 1954, pp. 372-274. 12

TABLE l: THE URANIUM SERIES Nuclide Half Life Radiation u238 4.49 x 109 y,7 92 $ Th234 24.1 d P 7 90 pa234 1.2 m i,7 91 Pa U234 2.48 x 105 y y 92 4 onTh230 8.0 x 104y,7 go Ra226 1620 y a,7 88 + 8Rn222 3.8 d c 86 84Po218 3.05 m 84 I r -| 0.04 pb214 V 26.8 m,r 82| 5At218 1.5 s 83Bi214 19.7 m a,^,r 0.04 4 84Po214 1.6 x 10-4 s a 81T1210 1.32 m P,7 82Pb210 22 y y,r.B210 4.99 d p 10-5 4Po210 138 d,7y 81T206 4.2 m 81 0 2Pb206 STABLE 13

TABLE 2 COMMON RADIOSOTOPES* Radionuclide Half life Beta particle, Mev Gamma ray, Mev H3 12 yr 0.018 None Li8 0.9 sec 12 Weak Be7 54.5 dayg K capture 0.48 (12%) Be10 2.5 x 10 yr.56 None B12 003 sec 13 Weak C14 5800 yr 0.155 None N13 10 min 1.24 (e+) None N 7.35 sec 10 (18%), 3.8-4.6 (82%) 6.2 015 2 min 1.68 (e+) None 019 29.4 sec 2.9 (70%), 4.5 (30%) 1.6 (70%) F20 12 sec 5.1 2.2 Na22 2.8 yr 0.575 (e+) 1.28 a24 15 hr 1.39 2.76 and 1.38 Mg27 29.6 min 1.8 (80%), 0.9 (20%) 1.01 (20%), 0.84 (100%) A2 2.3 min 3.01 1.8 i31 2.7 hr 1.6 None p32 14.3 days 1.71 None S35 87 days 0.167 None 136 4 x 105 yr 0.7 Weak C138 38 min 4.81 (53%), 2.77 (16%), 1.6 (31%), 2.15 (47%) 7 1-11 (31%) A37 34 days K capture, L capture None A41 1.8 hr 1.2 1.3 K42 12.4 hr 3.58 (75%), 2.04 (25%) 1.51 (25%) Ca45 152 days 0.25 None Sc46 85 days 1.49 (2%), 0.36 (98%) 1.12 (98%), 0.89 (100%) V52 3-9 min 2.3 1.45 Cr51 26.5 days K capture 0.32 (3%), 0.267 (weak) Mn54 310 days K capture 0.84 Mn56 2.6 hr 2.86 (60%), 1.05 (25%), o.845, 1.81 (25%), 2.13 0.73 (15%) (15%) Fe55 2.9 yr K capture None Fe59 47 days 0.46 (50%), 0.26 (50%) 1.3 (50%), 1.1 (50%) Co57 270 days 0.26 (e+) 0.131 Co60 5.3 yr 0.31 1.17 and 1.33 Ni63 85 yr 0.06 None Cu64 12.9 hr 0.57 (35%), 0.65 (e+' 1.34 (1%) 66 4.3m20%), K capture (45%) Cu^ 64.3 min 2.7 1.32 Zn65 250 days 0.32 (3% e+), K capture 1.11 (46%) (97%) Zn69 14 hr IT 0.439 Ga72 14 hr 3.17 max (see charts) 2.5 max (see charts) As76 27 hr 3.12 max (see charts) 2.1 max (see charts) As77 40 hr 0.7 None Se75 115 days K capture 0.405 max (see charts) 14

COMMON RADIOISOTOPES (Cont'd.) Radionulide Half life Beta particle, Mev Gamma ray, Mev Br82 36 hr 0.465 0.547, 0.787, 1.35 Br87 55.6 sec 2 (55%), 8 (45%), delayed 3 neutrons Rb86 19.5 days 1.82 (80%), 0.72 (20%) 1.1 (20%) 89 S5. 53 days 1.5 None YO Q61 hr 2.2 None Zr95 65 days 0.887 (2%), 0.4 (98%) 0.708 (98%) Nb95 35 days 0.146 0.758 Mo99 67 hr 1.2 (75%), 0.5 (25%) 0.141, 0.726 Tc99 3 x 105 yr 0.30 None Ru97 2.8 days K capture 0.23 Ru103 42 days 0.35 (50%), o.665 (50%) 0.5 (50%) Pd109 13 hr 0.95 None Ag110 270 days 2.86 max (see charts) 1.5 max (see charts) Ag 111 7 5 days 1.06 None Cd115 43 days 1.67 0.5 In114 50 days IT, 2.05 (97%), K capture 0.192, 0.715 (3%), 0.548 (3%) (3%) Sn113 112 days K capture 0.09 Sb124 60 days 2.37 max (see charts) 2.3 max (see charts) I131 8 days o.6o (85%), 0.32 (15%) 0.638 (15%), 0.364 (85%) 1135 6.7 hr 0.47 (35%), 1.0 (40%), 1.3, 1.7 1.4 (25%) I137 22 sec Delayed neutrons Xe135 9.2 hr 0.93 0.247 Cs134 2.3 yr 0.658 (74%), 0.09 (26%) 0.794, 0.602, 0.568 (26%) Cs137 37 yr 1.2 (5%), 0.51 (95%) 0.669 (from 2.6-min Bal') Ba131 12 days K capture 0.26, 0.5 (strong) Ba140 12.8 days 1.022 (60%), 0.48 (40%) 0.54 (40%) La140 40 hr 2.26 (10%), 1.67 (20%), 2.5 (6%), 1.6 (77%) 1.32 (70%) other low-energy gammas Ce141 28 days 0.56 (30%), 0.41 (70%) 0.141 (70%) Ce144 275 days 0.32 0.13 (strong) PrL42 19 hr 2.15 (96%), 0.64 (4%) 1.57 (4%) Pr143 13.8 days 0.92 None Nd147 11 days 0.78 (67%), 0.17 (33%) 0.035 (strong), 0.58 (weak) 147 2.7 yr 0.23 None Sm153 47 hr 0.8 (33%), 0.68 (67%) 0.10, 0.07 Hf181 46 days 0.42 0.34 (22%), 0.48 (78%) Ta182 122 days 0.50 1.2 max, many others w185 77 days 0.43 0.134 W187 25 hr 1.32 (30%), 0.63 (70%) 0.68 max, others Re 90 hr 1.09 (67%), 0.95 (30%) 0.132 (37%), 0.275 (23%) Os191 15 days 0.14 0.13, 0.04 Ir192 70 days 0.67 0.65 max, many others 198 0.411 Au19 2.7 days 0.97.411 15

COMMON RADIOISOTOPES (Cont'd ) RadioHRa - Half life Beta particle, Mev Gamma ray, Mev nuclide 197 Hg197 2.7 days K capture 0.077 Hg203 44 days 0.205 0.286 T120 2.7 yr 0.78 None Pb210 22 yr 0.028 Soft 210 Bi2 5 days 1.17 None po20 138 days 4.95 (alpha) None Rn222 3.82 days 5.49 (alpha) None Ra226 1620 yr 4.7 (alpha) 0.188 Th232 1.39 x 1010yr 4.1 (alpha) None Th233 23.5 min 1.2 None Th234 24.1 days 0.205 (80%), 0.11 (20%) 0.093 (20%) Pa233 27.4 days 0.58 max (see charts) 0.471 max (see charts) pa234 1.2 min 2.32 (98%), also IT See charts U233 1.6 x 105 yr 4.82 (alpha) 0.04 234 2.5 x 105 yr 4.76 (alpha) Weak 235 8.8 x 108 yr 4.5 (alpha) 0.17 236 2.5 x 107 yr 4.5 (alpha) None U238 4.5 x 109 yr 4.18 (alpha) None U239 23.5 min 1.2 0.074 uD240 14 hr Np239 2.3 days See charts See charts Pu239 2.4 x 104 yr 5.15 (alpha) Weak Pu240 6600 yr 5.1 (alpha) None *From Introduction to Nuclear Engineering, R. Stephenson, McGraw-Hill Book Co., Inc., New York, N. Y., 1954, p. 372. (IT) refers to isomeric transitions. (Charts) refers to more complete descriptions such as found in: Nuclear Data, National Bureau of Standards Circular 499; also, R. W. King, Reviews of Modern Physics 26, 327 (1954). 16

3.. NUCIEAR REACTIONS A. General In nuclear reactions, the total number of neutrons and protons is conserved and also the total charge is conserved. Thus in balancing an equation for a nuclear reaction, it is necessary that the sum of the mass numbers A for the reactants be equal to the sum of the A's for the products, and also that the sum of the charge numbers Z be the same for the reactants and products. In counting the charge numbers, any electrons involved must be counted also. For example: 9Pu39 92U235 + 2He Alpha decay of plutonium. 3 3 H 2He- + e-, Beta decay of tritium, 1 2 _Be9 + 2He4 -> 12 + nl, A neutron production reaction. U235 n+ n - e La147 + + Br87 + 2 n1 A fission reac92 0 57 35 o tion of uranium. 2 /-H 1 I D + X- H + on, Photo disintegration of deuterium. 1 1 Another way in which nuclear reactions are represented is by classifying them according to the bombarding particle and the prod.uct particle or the reaction process. For example: the third reaction in the above list is written 4Be9 (a,n)6C12 and is called an alpha,n reaction. 17

The deuteron photodisintegration is written 1D2(6,n)lH' and is called a gamma,n reaction. Neutron induced fission reactions are sometimes written as (n,f). Some other reactions often seen written in the literature in this fashion are the following: (n,2n); (n,p); (d,p); (d,p); (Yd); (d,np); etc. B. The Compound Nucleus Theory of Nuclear Reactions In the nuclear reactions involving medium to heavy weight nuclei in which free neutrons or protons are produced, the compound nucleus theory of Bohr has been very successful in explaining the experimental results and also provides a convenient physical picture. The compound nucleus is a highly excited nucleus which may have enough energy to emit a neutron, a proton, etc. The amount of excitation energy necessary to emit a single neutron or proton varies from one compound nucleus to the next, but on the average is about 8 Mev. According to Bohrts theory, the nuclear reaction may be thought of as occurring in three steps: 1. Excitation of the Initial Nucleus This may be accomplished by absorption by the initial nucleus of a neutron, proton, deutron, alpha particle, a gamma ray, etco 2. Formation of the Compound Nucleus In this stage, the energy of excitation becomes averaged in the nucleus with the result that one says the compound system has been formed. The process probably requires many nuclear periods to take place or a time of the order of 10-15 seco During this time, the compound nucleus forgets how it became excitedo 3. Decay of the Compound Nucleus If the excitation energy of the compound nucleus is greater than about 8 Mevo, a neutron or a proton may be emitted, A proton 18

is much less likely to be emitted than a neutron because a proton encounters an electrostatic barrier at the edge of the nucleus. Proton emission is likely to be less than neutron emission by a factor of 100 to 1000, and so in a nuclear reactor such protons may be neglected. If sufficient excitation energy is available, we may get two neutrons, a neutron and a proton, or a deuteron. For a given amount of excitation energy, the decay particles will be emitted in a spectrum with a maximum value corresponding to the excitation energy. The remaining energy is radiated as gamma rays. Some of the classes of reactions that may be described by the compound nucleus model are: (^,n); (O,p); (d,n); (n,2n); (~0,n), etc. C. Nuclear Fission Some of the very heavy nuclei are unstable against the breakup into two fragments. In a particular case, this may be determined by examining the binding energies of the fissioning nucleus and the prospective fission fragments. A few nuclei appear to undergo fission spontaneously, but in general the fission reaction requires some energy from an outside source to make the reaction proceed at an appreciable rate. This is similar to the situation with exothermic chemical reactions where heat or light is required to produce the activation energy. In the case of fission, the amount of activation energy required depends on the particular isotope involved. Some nuclei like U-233, U-235, and Pu-239 fission readily with the energy available from the capture of slow neutrons. On the other hand, U-238 requires at least 1.1 Mev. neutrons to cause the fission. Fission may also be induced by irradiation with high energy gamma rays. This latter process is referred to as photo-fission. The fission 19

process can be represented graphically as shown in Figures 4 and 5. In Figure 5, the ordinate is the potential energy of the system, and the abscissa is the distance of separation of the centers of gravity of the fission fragments. The amount of energy available by fission is Es, and the activation energy or critical energy is Ece FIGURE 4 THE FISSION OF A TYPICAL URANIUM-235 NUCLEUS 0 \Ba135 fission fragments incident neutron ~0~~ 0 U235 nucleus Key: * Proton 7 o Neutron Kr T~~ ~ Gamma Ray There is no unique mode of fission, i.e., fissioning nuclei may break up in many different ways to produce a variety fission product nuclides. The fission shown in Figure 4 is only one of these many fission modes. The fission fragments lie near the middle of the dotted line on Figure 1, drawn between the fissioning nucleus and the origin. It can be seen from Figure 1 that the fission fragments have a considerable excess of neutrons over stable nuclei with the same Z. As formed, the fission products are thought to be compound nuclei, as described above, which decay with the emission of one or more neutrons. The emission of only one or 20

co CO ww ZU Z U) z 2 m to~ ~ ~ ~' o ~c 0 i wU'.. ~~0ICO OW CO-)3N3 V~LN3~ aO. I rf) 31~~~~~~~~~~~~~U)U 0 U) y0000 z~~~~~~~~~ Lu_ z w~~~~~~~~~~~~~~~~~~~~~~~ Yb wO~~~~~~~ A98 ~b3N3 IV lILN3-Od U. Z r

two neutrons does not mean the fragments are stable, however, as they are still too rich in neutrons. Thus, these fragments undergo a series of beta decay processes in which Z increases and A remains constant until a stable nucleus results. A typical chain of beta decays is: z140 P, - 14 B a- 140 - La140 Xe T5Cs -;- Ba La -> 54 6 s 55 66 sec. 56 12.8 days 57 3- ~14o 40 hs. 58Ce (Stable) Gama rays are often radiated accompanying these beta decays, Some fission products are delayed neutron emitters, and are important in reactors since they augment the neutron flux available to cause fission. Delayed neutron emission by a radioactive nuclide occurs when there is a beta decay to yield an intermediate daughter nucleus in a highly excited state. If sufficient excitation is available, the intermediate compound nucleus then decays with the emission of a neutron in a time of the order of 10-15 seconds. We call the original parent nuclide a delayed neutron emitter since the neutron emission is delayed by the decay rate of the beta process which precedes it, Below in Table:3 is a list of half lives of delayed neutron emitters observed in the fission products of U235. TABLE 3 Fraction of Total Half Life, sec. Energy, Mev. Fission Neutrons 0 43.42.000ooo84 1.52 0.62.0024 4.51 0o. 43 o 0021 22.0 o 56.0017 55.6 0.25 o00026 22

The 55.6 sec. activity has been identified as being due to Br87, and the 22 sec. activity is associated with I137. The other activities have not been identified. In Figure 6 is shown the energy spectrum of prompt neutrons obtained from the fission of U235 and Pu239. Note that the form of these spectra is very closely given by Ce-E sinh(2E)1/2 where E is the neutron energy in Mev., and C is a constant. The fission product mass spectrum shows the probability of formation of a nuclide with mass number A when a heavy nucleus fissions. The probabilities are expressed in percent. Fission product mass spectra are shown in Figure 7 for the fission of U235, u239 and U238 by slow neutrons. Some data are also shown for the fast neutron fission of Pu239 It is of interest to have a breakdown of the energy evolved in the fission process. Recent estimates show that for the fissioning nucleus of either a U235 or Pu239 nucleus, the energy available is distributed in the following manner: Mev per Kilowatt hrs. X3 fission per fission. _ O Kinetic energy of fission fragments............. 162 ~ P Kinetic energy of prompt neutrons * H U produced in the fission process............... 6 $.~ Energy of instantaneous gamma rays............. 6 h. X Energy from absorption of excess neutrons rd i produced in the fission process which are e | a captured in non-fission processes by reactor XH X materials...................................... 8. X -y f Total 182 Mev 8.08 x 1018 C *m r X Energy from fission product gamma rays......... 5 X A.^ 0 a Energy from fession product beta particles..... 5 m -18 W Total Energy Available 192 Mev 8.52 x 10 0) Energy carried away by neutrinos accompanying - *~ a the fission product beta decays......... 11 Mev 0o*488 x 10 23

F/G. 6 SPECTRUM OF NEUTRONS PRODUCED BY SLOW FISSION OF U-235 AND Pu-239 WU U,eC E sinh(V2E) 2 AREA UNDER CURVE EQUALS THE NUMBER OF NEUTRONS PRODUCED PER FISSION (SEE TABLE 6) 0 1 2 3 4 5 6 7 8 NEUTRON ENERGY-MEV

F/G. 7 _____ _FISSION PRODUCT MASS SPECTRUM VI~ I r~~~I I......,,,,, 1~| -- |j,.- 1.0 P3_ _ _ _3Pu-239 1.0 2 1U-23 8 1 I / Pu-\9 4 2398 YIE!S INi' I__,N _ _ __I I I 1 I 1' \ I 8 I0 10 11 12 10 141 50 16 OFAST l In iM N A FISSION 1.1-235 " YIELDS IN I,MASS NUMBER A 80 90 100 110 120 50 140 150 160~I;,; MASS "UE l

The energy carried away by the neutrinos cannot be recovered since there is no known shielding material that will stop them. Although U235 is the most widely used nuclear reactor fuel, U233 and Pu239 can be used in certain reactors as well. Many other heavy nuclides fission besides these three. In order to be of practical use as fuel in nuclear reactors, a fissile nuclide must either be obtainable naturally or readily produced from naturally occurring materials. (See Part II, Section 2-D on fuel breeding). D. Fusion and Thermonuclear Reactions Fusion may be thought of as an inverse of the fission reaction. Thus fusion is a reaction in which two light nuclei combine to form a heavier nucleus. There are many fusion reactions which are exoergic, i.e*, will give up energy to the surroundings. For example: H1 + T3 > 2He + Q (energy) D2 + 5B10 6C12 + Q Even though these reactions are exoergic (Q is positive), they will not proceed spontaneously because the electrostatic repulsion between the two nuclei prevents their approach to a distance close enough for a reaction to take place. In order for these reactions to occur, the reactants must come together with a great deal of relative energy. This can be done either by using particle accelerators or very high temperatures. In the latter case, the reactions are called thermonuclear reactions, and it is believed such nuclear reactions are responsible for the large energy output from the sun and other stars.* * See for example, "The Birth and Death of the Sun," by George Gamow. 26

E. Nuclear Cross-Sections The rate at which externally induced nuclear reactions occur brings us to the definition of the cross-section. The reaction rate in nuclear events per cm3 of sample bombarded per second is proportional to the flux 0 of incident particles per cm2 per second inducing the reaction, and to the number N of nuclei per cm3 of the nuclear species of interest. (Reaction Rate) = -0 N The proportionality factor T is seen to have the dimensions of an area in 2 cm. ~ is called the cross-section for the nuclear event in question, and may be considered to be the effective area that the nucleus presents to the incident flux. Since many different kinds of nuclear events can take place, there are as many different kinds of cross-sections. In (n,7) reactions, neutrons are absorbed and gamma rays radiated, so the crosssection for these reactions is called the neutron radiative absorption crosssection 9c. In a similar manner, the fission cross-section Qf is defined, as well as others like the elastic scattering cross-section'sc which describes the rate at which an incident beam of particles is deflected due to elastic collisions with nuclei. Cross-sections are usually additive so that the rate, at which neutrons, for example, are removed from an incident beam is given by <f f (ac + 7f + q-sc + ) = No-^ where cft is called the total neutron cross-section. If the sample of material bombarded contains many different nuclear species, the rate of removal of neutrons from the incident beam is given by a sum of such terms. viz: 0 (Tl+ N22 + +- N+ nOn) 27

where N1 is the number of nuclei of species 1 per cm3 and 0t is the sum of the cross-sections for absorption and scattering of the incident particles, N2 and. q2 are the same for nuclear species 2, etc. For example, the rate of absorption of neutrons by natural uranium is given by: [N 23235( 53 + 35) + N238 8c -24 2 Cross-sections are usually expressed in units of 10 cm2 This unit is called the barn (b) and is of the same order of magnitude as the geometrical area that a nucleus presents to the beam of incident particles. Some very small cross-sections are given in units of 10-27 cm2 or millibarns (mb). r is often called the microscopic cross-section as it describes the reaction rate due to a single nucleus. The macroscopic cross-section on the other hand is defined as the product NHrwhere N is the number of nuclei of the type under consideration per cubic centimeter of material. The units for are cm1. Neutron cross-sections are dependent on the energy of the incident neutron. For very low energy neutrons or thermal neutrons.,* most neutron absorption cross-sections are proportional to 1/v where v is the velocity of the incident neutron. This is often called the 1/v law. Extensive tables of thermal neutron cross-sections are given in an unclassified document.** Thermal neutron absorption cross-sections range from a fraction of a millibarn to over a million barns, while nuclear areas are all of the order of * Thermal neutrons have energies in the range between zero and 0.1 electron volts. They are called thermal neutrons because a neutron gas in thermal equilibrium with its surroundings at room temperature has a distribution of energies in this range. * Neutron Cross-Sections, AECU-2040, obtainable from the Office of Technical Services, Department of Commerce, Washington 25, D. C. 28

one barn, and so there is little correlation between the neutron absorption cross-sections and nuclear size at thermal neutron energies. At high neutron energies, (greater than about 15 Mev) the total neutron cross-sections approach the value 2 TRB where R is the nuclear radius. Thermal neutron cross-sections are usually stated at an incident. neutron energy of 0.026 electron volts which means the neutrons have a velocity of about 2200 meters per second. (See Table 4) Of particular interest in reactor design are the thermal neutron capture and fission cross-sections for the fissionable fuel metals. A table of the values of some of these cross-stions is givensectionis given in Part I, section II-A. Plots of total neutron cross-sections versus neutron energy for some elements important in reactor design are shown in Figure 8. This velocity of 2200 meters per second = 7,219 feet per second 4,922 mph is chosen as it corresponds to the most probable neutron velocity found in a neutron gas in thermal equilibrium at 27~ C. (81~F). 29

TABLE 4. THERMAL NEUTRON ABSORPTION CROSS SECTIONS* isotopic Element Isotope Abundance, Cross Section, per cent barns H -- 0.33 H1 100 0.33 H2 0.0150.46 mb He — Variable He 0.00013 np 5200 He 100 0 Li -- -67 Li6 7.5 na 910 Li7 92.5 33 mb Be Be9 100 9.0 mb B --- 750 B10 18.8 na 3990 B11 81.2 50 mb C -- 4.5 mb C12 98.9 C13 1.1 1.0 mb N. —1.78 N14 99.6 np 1.70, ny0.10 N15 0.37 0.024 mb 0 - -- 0.2 mb 06 99.76 Very small 017 0.037 na 0.5 018 0.20 0.21 mb F F19 100 10 mb Ne ---- 2.8 Na Na23 100 0.49 Mg --- 59 mb Al A127 100 0.22 Si -- 0.13 P p31 100 0.19 S -- - 0.49 C1 -- -- 31.6 A ---- 0.62 K -- -- 197 Ca -— 0.43 Ti ---- 5.6 V -- -- 4.7 Cr -- 2.9 Mn Mn55 100 12.6 Fe - 2.43 Co Co59 100 34 Ni ---- 4.5 Cu - — 3.59 Zn -- -- 1.06 Zr ---- 0.18 Mo ---- 2.4 30

THERMAL NEUTRON ABSORPTION CROSS SECTIONS (Cont'd.) isotopic Element Isotope Abundance, Cross Section, per cent barns Cd - - 2400 In - -- 190 Sn -- - 0.65 Xe.35 e35 0 35 x 10 Sm - -- 6500 Sm149 13.8 50,000 Eu - -- 4500 Gd -- - 44,000 Hf - -- 115 Ta - - 21.3 Au Au197 100 94 Hg - -- 380 Pb 0.17 Bi Bi209 100 32 mb Th Th232 100 7.0 Th233 0 1400 Pa Pa233 0 37 U - - nY 3.50, nf 3.92 U23O 0.714 nr 101, nf 549 U235 99*3 2.80 U239 o 22 Pu Pu239 0 n 361, nf 664 *From "Introduction to Nuclear Engineering," R. Stephenson, McGraw-Hill Book Co., Inc., New York, N. Y., 1954, p. 375. mb means "millibarns" or 10-3 barns. -27 2 One mb = 10 cm 31

0 Q C.) z to I _ _ _ 0 wI z z,:: w I.0 z i, 0 -o z, z TFFT 1 z 0(O CV WW 4 UIfl 0.L.:/~~~ o o o o o~.~~~~ co n,"c lu ~ ~nn ~ ~ ~ 1

PART II - NUCLEAR CHAIN REACTION - REACTOR MATERIALS.1CHAIN REACTIONS AND NUCLEAR REACTORS - GENERAL DISCJSSION A. Nuclear Chain Reaction; the Critical Fuel Mass On the average, 2~5 neutrons are produced in the fission of one U235 nucleus. These neutrons can cause fission of other U235 nuclei and so in principle, it is possible that a chain reaction can occur in which new fissions are being caused by neutrons produced in previous fissions. These considerations also apply to other fissionable materials such as U233 and Pu239o A nuclear reactor is an assemblage of fissionable fuel, structural and other materials in which a nuclear chain reaction can take place. (A specific type of nuclear reactor of historical importance is the pile, which consists of a matrix of metallic uranium slugs and blocks of graphite,) The atomic bomb is a nuclear reactor which is specifically designed to momentarily contain an uncontrolled chain reaction until the fuel is almost completely consumed, before destroying itself. Nuclear reactors in which the chain reaction is controlled are used to produce intense neutron beams for physics research, to produce radnioactive nuclel of species not obtainable naturally for medical and radioactive tracer studies, to manufacture plutonium from uranium for use-in the weapons program, and to produce heat for power generation, ship propulsion, etc. Of the neutrons produced in a chain reaction, many either escape the region in which the fuel is concentrated, or are captured in (n,y) reac238 tions with U238 and other nuclides present. In order for a chain reaction to be self-sustaining, the number of neutrons producing new fissions must be equal to or greater than the number of fissions producing themo Another 33

way of saying this is that the number of neutrons in one generation should be equal to or greater than the number in the previous generation, or that the ratio of these numbers be equal to or greater than unity. This ratio is called the multiplication factor k. k = (No. of neutrons in one generation) No. of neutrons in the previous generation) To be more precise, it is the product k P = k, the effective multieff plication factor, which must be greater than or equal to unity to insure a chain reaction. Here P is the probability that neutrons do not escape from the edge of the reactor, or the non-leakage probability. A reactor having keff < 1 is said to be sub-critical and a chain reaction cannot be maintained. When keff = 1, the reactor is critical and the reaction is self-sustaining. If keff > 1, the reactor is super-critical, and the reaction tends to build up. The multiplication factor k, defined above, is dependent on microscopic quantities such as the fission and capture cross sections. On the other hand, the non-leakage probability P is dependent mainly on the size and shape of the whole reactor. If the reactor is small, there is a good chance that neutrons produced in the reactor will escape without causing fission and therefore P is small. For a given k and geometrical form.for the reactor, there is a minimum reactor size for which kP = 1. This is the critical size of the reactor. The mass of fuel in the reactor of critical size is the critical mass. It should be noted that for a given reactor volume, neutrons more readily escape from a reactor with a large surface area, and so the critical mass may be kept at a minimum if the area is a minimum or the reactor is spherical. Another way of keeping the critical 34

mass small is to use a moderator, and take advantage of the large thermal neutron fission cross sections of fuels. These large cross sections reduce the average distance a neutron travels before being captured by a fuel nucleus, and consequently reduce the leakage. It is necessary'to keep in mind the possibility of inadvertently assembling a mass of fuel as metal or in solution that will be of critical mass. This caution is important mainly in dealing with uranium enriched in U235 or with Pu239. For example, about 800 grams of U235 in aqueous solution contained in a spherical vessel of one foot in diameter can be critical if a neutron reflector is brought near. A technique of experimentally determining the critical size and critical mass of an assembly of fuel and moderator is the construction of the critical assembly. Here an external neutron source such as the radiumberyllium source is placed in the center of a small amount of the fuelmoderator assemblyo The neutron flux is measured by two or more neutron detectors placed in various locations, and the fuel-moderator combination is added in small amounts. A plot of the reciprocal of the measured counting rates from the neutron counters vs. the total fuel mass is kept concurrently which will be of the form shown in Figure 9, When the critical mass is reached, the multiplication of the neutron flux from the Ra-Be source by the assembly will be infinite or the reciprocal counting rates will be zero. However, it is not necessary to add enough fuel to make the assembly critical, since an extrapolation of the reciprocal counting rates to zero will intersect the horizontal axis at the critical mass. The alpha particles from the decay of radium react with beryllium to give neutrons through the reaction: Be (a,n)C12. 35

T FIG. 9 wF^ -~~DETERMINATION OF THE /:::S:-'CRITICAL MASS BY THE CRITICAL ASSEMBLY EXPERIMENT Z0 IrL 0 CRmCAL S^^<.o~~~~ X \ MASS 0 250g 500g 750g 10OOg 1250g FUEL MASS --

Anything that captures neutrons, other than material which will convert to fissionable materials, may be termed a reactor poison. This includes the fission products that are formed as a result of fissioning, the control rods that are purposely introduced to absorb neutrons, and structural materials which may be inherent in fuel element design or structural components of the reactor itself. B. Reactor Classification Based on Neutron Energies The fission cross-section for natural uranium is about one barn for neutrons with energies in the neighborhood of 1 Mev. For thermal neutrons, however, the fission cross-section for U235 is 549 barns (See Table 63), and so it may be advantageous to have thermal neutrons available to cause new fissions. Since the neutrons produced by the fission process have average energies of about 1 Mev (See Figure 6), they must be slowed down by substances called moderators to thermal energies in order to take advantage of the high thermal neutron fission cross-sections of U235 and Pu239. Moderator nuclei slow neutrons by absorbing their energies in recoils from elastic collisions. A good moderator has a small neutron capture cross-section and a small mass number A. Moderators are discussed in more detail in section III-B. Reactors using thermal neutrons to cause fission are called thermal reactors, and those using fast neutrons (no moderator used) are called fast reactors. C. Reactor Classification Based on Fuel Distribution Reactors may be further classified according to the distribution of fuel. If the fuel is distributed uniformly as in the case of an aqueous solution of a uranium salt or a molten solution of uranium metal in another 37

metal, the reactor is called a homogeneous reactor. If on the other hand, the fuel is lumped into fuel elements separated from one another by the moderator or some other substance, we have a heterogeneous reactor. D. Reflectors and Blankets In order to reduce the escape of neutrons from a reactor, a reflector is used. The reflector action is diffuse reflection and is due to the neutrons being elastically scattered by the nuclei of the reflector material back into the reactor core. Good reflector materials have large neutron scattering cross-sections and small capture cross-sections. Very pure graphite is often used. A beam of neutrons incident on a 15-inch thick slab of graphite is 90% diffusely reflected, i.e., the slab has an albedo of 0.90. Sometimes a natural uranium blanket surrounds the reactor core to capture neutrons which would otherwise be lost. Capture by U238 yields U239 which decays by two beta emissions to Pu239. In this way, fissionable Pu239 may be made from non-fissionable U238. Similarly, a thorium blanket can be used to produce U233 which is fissionable. This is referred to as fuel breeding, and is discussed in more detail in section II-E. (See Fig. 10.) 2,. NUCLEAR FUELS A. Natural Uranium The main mineral sources of uranium in the world are shown in Table 5. The pitchblende are the richest uranium ores and consist of U308 and ores of other metals such as lead, copper, and silver. Carnotite is a uranium-vanadium mineral, K20.2U03.VT205.niO2, and is found in Colorado and Utah as a cement in certain sandstones. Carnotite is often found in combination with other uranium-vanadium minerals. 38

~~~~ 2~~~ cml ~ C8,,..X "~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~'~ IC~~~~~~~~~~~~~~~~~~~... B...<~ i~.~..!.::...Lr~.::::.:.:.:::.~~.:..;. Q,.,,..,..,,.,,Z 5~3~~~~~~~~''!,' 35g~~~~~,L,

TABLE 5 SOURCES OF URANIUM APPROX. RESERVES, MATERIAL SOURCE % U TONS URANIUM* Belgian Congo 5-50 Pitchblende Canada \ 1 20,000 Czechoslovakia 0.2-2 9 Carnotite f Siberia ~ Colorado 0*2-1 35,000 Gold Ore South Africa >0.01 Phosphate Rock Florida / 0.01 Sweden 0.025 Bituminous Shale Russia 0.01-0.025 2,000,000 Eastern U. So A. 0.003-0.01 1 *. George W. Bain, Geology of the Fissionable Materials, Econ. Geol. 45, 274 (1950). 40

The process of refining pitchblende from the Belgian Congo and Canada yields NaU207 or (NH4)2U07, the latter of which may be ignited at 1000~C to give U308. Carnotite ores are more difficult to treat* than pitchblende since the low uranium concentration necessitates handling large quantities of ore to get a given yield of Na2U207 or U308. Metallic uranium may be produced from uranium compounds in one of the following ways: a) Reduction of uranium oxides with carbon in an electric arc furnace. b) Reduction of uranium oxides with aluminum, magnesium, calcium, or calcium hydride. c) Reduction of uranium halides with alkali or alkaline-earth metals. d) Electrolytic reduction of uranium halides. e) Thermal dissociation of uranium iodide. (hot wire method.) 1. Physical Properties of Uranium Metal Density 19.050 g/cc Melting point 1133 + 2~C 2051 i 4~F Estimated boiling point 4470~C 8100~F Estimated heat of fusion 10 to 13 cal/gram For a detailed account, see "Studies of Recovery Processes for Western Uranium-Bearing Ores," V. L. Saline and K. B. Brown, AECD-3241, October, 1949. 41

The breaks in the specific heat and heat content curves of Figure 11 are evidence of solid state phase changes. The low temperature phase is alpha uranium, the next is the beta phase, and the high temperature phase is the gamma phase which exists up to the melting point. The temperatures at which the phase changes occur are measured by many different methods, and have some dependence on the history of the metal. Typical resistivity data yield: Uranium Metal Phase Change Temperatures Heating Cooling a ~ > 0 P 7 7 e r > 0 P a T~C 667 772 764 645 T0F. 1233 1422 1407 1193 The electrical conductivity of uranium metal at room temperature is about 2 x 104 to 4 x 104 (ohm - cm)-, or about half the conductivity of iron. The thermal conductivity at room temperature is about 0.06 cal/cm"sec-~C. At 300~C (572~F), it is 0.075 cal/cm-sec-~C. The tensile strength varies between 50,000 and 200,000 psi. The variation depends on the history of the metal. Uranium may be cold worked to give high tensile strengths or annealed to give lower values. The tensile strength decreases to 12,000 psi at 600~C (ll12~F). 2. Chemical Properties Uranium metal is highly reactive chemically. It forms hydrides with hydrogen gas at 2500C, and burns brightly in oxygen at about 7000C (1292~F). Carbides, nitrides halides, and other compounds are also readily formed. Metallic uranium also reacts with steam and the mineral 42

SPECIFIC HEAT OF URANIUM AT ELEVATED TEMPERATURES I J-12 — o1I —- --- -- I — - i -- i - i - - i - i, - o 13.000 11.0 X' 11,000 r / *- 2 < FIG. I 10.0 9,000, 9.0 / {oo 7,000 8.0 5,000 U 4 0 - o:f f' Io 7.0 3P00 /,'- SPECIFIC HEAT d,/S. -—...HEAT CONTENT 6.0 -- 1,000 200 400 600 800 1000 1200 1400 TEMPERATURE,~K From "The Chemistry of Uranium',' Katz & Rabinowitch, Mc Grw - Hill Book Co. 1951.

acids. The oxidation potential of uranium is believed to be close to that of beryllium. Uranium is a strong reducing agent in aqueous solutions. 3. Nuclear Properties The radius of the uranium nucleus is about 8.7 x 10-13 cm. For this radius, the cross-sectional area of a uranium nucleus is 2.4 x 10-24 cm2 or 2.4 barns. It is interesting to note that the neutron fission and capture cross sections given in Table 6- are much larger than 2.4 barns. Explanation of the anomalous result is based on wave mechanics or quantum mechanics in which particles are represented as waves. A situation analogous to neutron capture by a uranium nucleus is a sound wave incident on a small region that strongly absorbs sound waves. In this analogous problem, the absorbent region appears larger by sound absorption measurements than its physical dimensions. A list of the uranium isotopes, their half lives, and natural abundances are shown in Table 7-* The cross sections for neutron capture and fission of U235 are shown in Table 6 TABLE 6 THERMAL NEUTRON FUEL METALS CROSS SECTIONS Nuclide U235 U238 Pu239 Th232 Cross Section, barns fission 549 -- 664 neutron capture 101 2.8 361 7.0 + 0.4 Total absorption 650 2.8 1025 7.0 + 0.4 Average number of neutrons produced per fission 2.5 + 0.1 -- 30 0.1 44

B. Thorium Although natural thorium, which is very nearly 100% Th232, does not fission with thermal neutrons, it captures neutrons to form Th233, which is beta radioactive with a half-life of 23*3 min. decaying to protactinium-233 (Pa233). Next the Pa233 beta decays to U233 with a halflife of 27.4 days. Uranium-233 fissions with slow neutrons and so natural thorium is a potential fuel material. It is estimated that there is about three times as much thorium on the earth's surface as there is natural uranium. The principal source of thorium is presently the monazite sands found in India and Brazil. Monazite is a rare-earth phosphate mixture containing a few percent of thoria. (thorium dioxide, ThO2.) The monazite is treated mechanically to increase the thoria concentration. It is then dissolved in sulfuric acid and after a long series of operations necessary to remove the rareearths, thoria is obtained. Thorium metal may then be obtained by: a) reduction of ThC14 by sodium, b) reduction of thoria by calcium metal, and c) electrolysis of KThF in a bath of molten NaCl-KC1. 5 1. Physical Properties of Thorium Metal Specific gravity 11.7 Melting point 15000C appx. 200 2700O~F Boiling point ap. 000~ 7200~F Thermal conductivity at 1000C 0.076 cal/sec-cm-~C Electrical conductivity at 20~C Appx. 4 x 104 (ohm-cm)-1 Specific heat Cp = 0.03437 + 0.198156 x 10-5 T + 0.43152 x 10-8 T2 + 0.452056 x 10-11 T3 Cal/~C 45

2. Nuclear Properties The thorium nucleus is about two percent smaller in size than the uranium nucleus. The thermal neutron capture cross section is 7.0 ~ 0.4 barns. The thorium isotopes are listed in Table 8: C. The TlaansraaiumI Elements - Plutonium The transuranium elements are those which occur as a result of the neutron irradiation of uranium. A schematic showing some of the reactions leading to the formation of some of these nuclides is shown in Figure 12. Not shown are elements 97 (Berklium, Bk) and 98 (Californium, Cf). Of these elements, plutonium has received the most attention in that it fissions under irradiation with thermal neutrons and is readily formed by irradiation of the non-fissile U238. Large quantities of Pu239 are produced in water-cooled graphite moderated reactors located at Hanford, Washington. Aluminum clad uranium slugs are inserted in channels through.the graphite and after a given exposure to the reactor neutron flux, are pushed out the back of the reactor by the insertion of new slugs. The exposed slugs are then treated chemically to obtain a separation of the plutonium produced from the remaining uranium and the fission product elements. This chemical separation must be performed by remote control because of the intense radioactivity of the fission products. The nuclear reactions leading to Pu239 from U238 may be traced on Figure 12, and are specifically written down in section II-E. The isotopes of plutonium and their modes of radioactive decay are listed in Table 9* The thermal neutron fission and capture cross for Pu239 are given in Table 6. This plutonium plant was built during World War II as an important part of the weapons program.'6

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More detailed information on the physical, chemical, and nuclear properties of the nuclear fuels may be found elsewhere..* D. Fuel Enrichment Reactors for ship propulsion, as well as for other mobile nuclear power plants, must be as physically compact as possibleo This necessitates the use of fuels rich in U235 or Pu239 (and perhaps in the future, U233) since for enriched fuels, the critical fuel mass may be made quite small. Enriched fuel reactors in operation at present mainly use U235, said to be worth about $20O00 per gram, rather than Pu239, which is reported to be several times more expensive. Since 1944, a large scale plant involved in the separation of U235 from natural uranium has been in operation at Oak Ridge, Tennessee. The plant output is mainly used in the weapons program. The process used is the gaseous diffusion process which makes use of the slightly different permeability of thin membranes to UF6 (uranium hexafluoride) gas, depending on whether a U235 or U238 atom is in the UF6 molecule involved. UF6 gas is introduced into a diffusion cell with a membrane stretched across it. On the other side of the membrane, the pressure is somewhat lower and the gas diffuses through the membrane; a slight enrichment in U235 taking place in the process. A large number of diffusion cells are necessary to effect an appreciable enrichmentO The cells are cascaded in the form shown in Figure 13. At the level of high enrichment, fewer cells are necessary as much less material is being handled. The entire plant at Oak Ridge is extremely large and many new techniques and materials had to G. T. Seaborg and J. J. Katz, The Actinide Elements, National Nuclear Energy Series, McGraw-Hill Book Co., Inco, New York N.Y., 1954; also, J, J, Katz and E. Rabinowitch, The Chemistry of Uranium, N. N, Eo S., 1951. 49

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be developed in order to make the project an operating reality. Other gaseous diffusion plants are located at Paducah, Kentucky and Portsmouth, Ohio. An attempt was made to develop a practical means of separating U235 and U238 on the basis of mass spectrography. This involves electrically accelerating uranium ions inside a vacuum chamber and allowing them to drift through a region in which an intense magnetic field has been produced as in Figure 14. The magnetic field causes the ions to follow circular paths where the radius of the circle depends on the mass of the ion. In this way, an excellent separation may be obtained, but the amount of material that can be handled is very small and so this technique is used only in thr production of small quantities of certain isotopes for research. This separation process is called the electromagnetic separation. E. Fuel Breeding Two fissile nuclides which do not occur naturally to an appreciable extent are Pu239 and U233. These may be obtained by neutron irradiation of U238 and Th232 as follows: U238(n,y)U239 -p Np239 B- Pu239 23.5 min. 2.33 days Th232(n,7)Th233 *- pa233 B- U33 23.3 min, 27.4 days It is possible in principle to produce more fissile fuel than is consumed in reactors, although in practice this is difficult of attainment because many reactor neutrons are absorbed in (n,r) reactions with poisons such as the fission product nuclides, structural materials, etc. 51

The reason that breeding is possible is that more than one neutron is produced in fission for each one absorbed by a fissioning nuclide. For example, if natural uranium is used as a fuel in a thermal breeder reactor, the fissioning of U235 yields 2.5 neutrons. Of these, a certain number are absorbed by other U235 nuclei in radiative capture,. (i.e., in the reaction U235(n,y)U236.) The percentage remaining available for other fissions and breeding is seen to be gO;_ = 549 barns = 84.5%, g + 101 barns + 1549 barns or (.845) x (2.5) = 2.11 neutrons remaining. One of these neutrons is necessary to keep the chain reaction going and some of the remaining 1.1 neutrons may be captured by U238 nuclei to give U239, which becomes Pu239 according to the above equation, A breeder reactor is usually defined as a reactor producing more fissile fuel than it consumes, and a converter reactor produces somewhat less fuel than it consumes. Another system of nomenclature defines a breeder as a reactor that burns the same nuclide that it produces, and a converter as a reactor that burns one nuclide and produces another. In this latter nomenclature, no reference is made to the amount of fuel produced and it is necessary to state if the fuel gain factor is positive or negative. (For example, see Table 10). It has been shown with the Experimental Breeder Reactor (EBR) at Arco, Idaho, that system 8 of Table 0 has a positive gain factor, and that the breeding principle is in fact realizable for a fast reactor. Table 10 shows many other possible fuel production schemes with predicted gain factors. 52

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Production of Pu239 or U233 can take place in fuel elements within a reactor or in a blanket of natural uranium or thorium. The build-up of Pu239 or U233 in a sample of potential fuel material follows a curve of the form shown in Figure 15. The equilibrium value of the Pu239 or U233 concentration occurs when that rate of production is equal to the fission rate. Also shown on Figure 15 is the build-up of fission product poisons. After sometime, it will no longer be economical to continue exposure of the sample to the neutron flux because of the loss in neutron economy due to the presence of the fission product poisons. There comes the problem then of performing a chemical separation in which fission products are removed. It may also be advantageous to separate the new U233 or Pu239 fuel that has been produced from the thorium or natural uranium. Plutonium is more difficult to separate chemically from uranium than is thorium, and so it may be economically advantageous to use the thoriumuranium breeding cycle than the uranium-plutonium cycle. Plutonium is also very posionous. These chemical separations must be carried out by remote control as the fission products are intensely radioactive. It is necessary to provide a thick concrete shield around fission product separation plants. Operation may be viewed with periscopes, through thick lead glass windows (specific gravity -6.2), or through glass walled cells containing aqueous solutions of zinc bromide (specific gravity -2.5). A variety of remote controlled manipulators are available which permit handling of chemical equipment, metal ingots, etc. from behind shielding walls. 3a 1REACTOR CONTROL AND REACTOR MATERIALS A. Reactor Control In a reactor operating in a steady state condition, the effective multiplication factor kff is exactly unity, regardless of the 54

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power level at which the reactor is operating. Reactor controls are needed to maintain the power level at a given value, and to make it possible to start the reactor and shut it dovwn In discussing reactor dynamics and control, it is helpful to introduce the term reactivity. The reactivity p is defined as: e (kef -l)/keff The numerator is seen to be the neutron multiplication in excess of what is needed to maintain the chain reaction and so, if the reactivity is positive, the reaction builds up, if negative, the reaction dies out, and if 0 = 0, the reactor is at a steady state. Factors tending to reduce the reactivity are: fuel depletion due to burn-up, the presence of poisons, such as structural materials, coolant, and moderator, the build-up of fission product poisons, the intentional introduction of control rod poisons, and the removal of either fuel or the neutron reflector. Reactivity has a temperature dependence because the thermal expansion of the reactor materials allows the escape of neutrons and also because at higher neutron temperatures, the absorption and fission cross sections are lower, In general, a reactor tends to shut itself down after continued operation mainly because of the accumulation of fission products. Samarium149 and xenon135 are the most serious fission product poisons encountered. Sm149 is a stable nuclide which results from the beta decay of Nd149 through: Nd149.- Pm1l49 - Sml49. (Stable) 1.7h 47h The fission yield of Nd149 is 1.4%. The thermal neutron absorption cross section of Sm149 is 5 x 104 barns. Xe135 is an intermediate in the decay chain: Te135 - I135 g- Xe135 - Cs135 - Ba135 a2m 6.7h 9.2h 2x10 y (Stable) The fission yield of Te135 is 5.6% in the thermal fission of U235. The thermal neutron absorption cross section of Xe135 is 3.5 x 10 barns and so it is the most prominent poison produced. 56

Thermal reactors may be controlled with rods or plates of neutron absorbing cadmium or boron steel. These may be moved in and out of the reactor by the use of servo operated motors which can be controlled by a neutron flux meter or a temperature indicator and in this way, the flux or power level may be kept constant automatically. Manually operated control of the rods can provide for reactor start-up and shut-doawn operations. Boron10 and cadmium l3 are used in control rods because these nuclides have high thermal neutron capture cross sections. For fast neutrons, however, these nuclides have small capture cross sections and cannot provide control of a fast reactor. Fast reactors may be controlled by fuel motion or reflector motion. Reflector motion is more difficult because of the large mass of materials that must be moved, but can be used in emergency scram control. It might be expected that the reactor control mechanism must be capable of extremely rapid response, as the time interval between two successive generations of neutrons is of the order of a millisecond in thermal reactors and even less in fast reactors. Fortunately, about 0.75% of the neutrons resulting from the fission of a fuel nucleus are not emitted immediately as are the rest, but are delayed by an average time of about 0.1 sec. (See Part I, Sec. 3-C for more information on delayed neutrons.) This time delay means that reactor control can be effected with slower control mechanisms if positive reactivity changes are kept less than 0.75%. If a reactor becomes critical on prompt neutrons alone (keff = 1.0075), it is said to be prompt critical, and slowly operating control mechanisms are of little use in stopping the build-up of the reaction. 57

B. Moderators Thermal reactors use very low energy neutrons or thermal neutrons (average speed 7200 ft./sec.) to cause fission of the fuel. Neutrons resulting from fission, however, have much higher energies with an average value of one Mev. (average speed 45 million ft./sec., see Figure 6 for the fission neutron energy spectrum), and must be slowed down to thermal energies by substances called moderators. The neutrons are slowed down by a series of elastic collisions with moderator nuclei. Figure 16 shows a typical neutron-moderator collision. The average percentage energy loss per collision is given by 2A/(l + A), where A is the mass number of the moderator nucleus. This function has a maximum at A = 1, and so hydrogen (A = 1) in some form is a good moderator. Other light weight nuclei given in order of increasing A and therefore decreasing moderating efficiency are: deuterium, helium, lithium, beryllium, boron, carbon, nitrogen, and oxygen. Of these, lithium and boron are unsatisfactory as they have large neutron capture cross sections and act as poisons, Moderators in the gaseous state are inefficient, as gases have such low density that the mean free path of a neutron between collisions with moderator nuclei is so long that a gas moderated reactor would have to be very large. The ideal moderator is dense, has a large neutron scattering cross section, small neutron capture cross section and small mass number. Some moderators and their properties are listed below in Table 11. Heavy water (D20) is composed of heavy hydrogen (deuterium) and oxygen. Heavy water is a very good moderator and may be used as a reactor coolant as well. At $20 per pound, heavy water might well find use in 58

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TABLE:1i PROPERTIES OF MODERATORS Material HO D20 Be C BeO 2 2 Density 1.00 1.10 1.84 1.6o 2.80 Atomic or molecular weight 18 20 9 12 25 Atoms/cm3 or mole- 2 cules/cm3 3.35xlO22 3.32x1022 1.23x10 23 8.05x1022 6.75x1022 at 0.025 ev, barns o.66 0.92 mb 9 mb 4.5 mb 9.2 mb Cs at 0.025 ev, barns 110 15 6.9 4.8 11.1 Epithermal Is, barns 46 10.5 6 4.8 9.8 Moderating ratio 67 5820 160 169 180 Slowing down length, cm 5.7 11.0 9.9 18.7 12.0 Slowing down time, sec 10 5 4.6xlO-5 6.7x105 1.5x10-4 7.8x10-5 Albedo (infinite) 0.82 0.97 0.89 0.93 0.93 60

the competitive nuclear power field. Heavy hydrogen occurs naturally with an isotopic abundance of 0.0oi%, and may be separated from natural hydrogen by either distillation processes or chemical exchange reactions. Mass spectrography is not a practical method since such large quantities of material must be handled to get an appreciable yield of deuterium* Of the possible processes listed in Table 12, the fractional distillation of liquid hydrogen at -252.70C offers great promise since the vapor pressure ratio is high. If there is sufficient external demand for liquid hydrogen, utilization of this process could bring the price of heavy water down to a reasonable value. The two other distillation processes in Table 12 require many separation stages to yield an appreciably enriched product. The steam distillation requires an economical source of heat. Consideration is being made of using the heat from natural hot springs in New Zealand in this process. The first of the exchange reactions listed in Table 12 is used at Trail, British Columbia and in Norway in the gas phase. The large amounts of hydrogen gas needed in the separation is eventually used in the synthesis of ammonia. The second of the exchange reactions of Table 12 using hydrochloric acid has promise of becoming an economical method as it can be carried out in the liquid phase without catalysis. Ordinary water is being used as a moderator in many reactors. It is necessary to use enriched fuels in water moderated reactors because many neutrons are captured by the hydrogen in the water. Swimming pool reactors and water boiler reactors use ordinary water as a moderator. Graphite is employed in many thermal reactors today. It is necessary to obtain very pure graphite for this service. The presence of 61

TABLE 12 DEUTERIUM SEPARATION PROCESSES DISTILLATION N.B.PT., C. VAPOR PRESSURE RATIO H20 - HDO 100 1.017 NH3 - NH2D -33.4 1.037 H2 - HD -252.7 1.73 EXCHANGE REACTIONS EQUIB. CONST. AT 25~C H20 + HD =- HDO + H 3.70 H20 + DC1 a: HDO + HC1 4.69 62

even small traces of boron is not allowable because of the large neutron capture cross section of boron. This impurity may be removed by passing freon gas through hot graphite* Beryllium as BeO is a good moderator material, but presesnt costs are high. BeO can be pressed to form bricks with a specific gravity of 2.8. Beryllium oxide is used as a reflector material in the Los Alamos water boiler reactor. C. Coolants The problem of removing heat from a reactor has resulted in the development of many new materials, mechanisms and techniques..Most of the reactors in operation today are water cooled and a few are air cooled, but liquid metal coolants are viewed as being much more satisfactory for the removal of heat from large power reactors now being planned. In reactors built to produce neutron beams for physics research or radioisotope production, the heating of the reactor core is in general a nuisance and operation at a low temperature is usually desired as it simplifies the design and operation considerably. Where the reactor is to be used in electric power production using a heat engine, high operating temperatures are preferable to give high thermodynamic efficiencies. Extracting reactor heat poses problems different from those encountered in coal or oil burning power plants. The heat produced per unit volume in a reactor can be made very high to keep the fuel inventory and shielding requirements low with the result that the heat transfer surfaces are small and require high heat transfer coefficients. Further, the primary coolant must have a small neutron capture cross section. Liquid sodium is presently receiving a great deal of attention and most proposed 63

power reactors plan on using sodium as a coolant, as it satisfies many of these requirements. In this plan, sodium is pumped at high velocity through the reactor core to a heat exchanger where the heat is removed by steam or by some intermediate coolant. An electromagnetic pump has been developed which requires no seals or moving parts, and considerably simplifies handling a liquid metal. These pumps operate by passing an electric current across the pipe containing the liquid metal. A force is exerted on the liquid metal by the interaction of a strong megnetic field which is arranged so that it is mutually perpendicular to the direction of the axis of the pipe and the direction of the electric current. Resistive losses in the liquid metal and the magnet coils reduce the efficiency of these pumps considerably. They operate best on direct current. Another type is designed to use alternating current, if a further reduction in efficiency can be tolerated. An alloy of sodium and potassium which melts at room temperature is another candidate for coolant service. It has the disadvantage of a higher neutron capture cross section than pure sodium, but otherwise is quite satisfactory. For high temperature homogeneous reaction, a solution of uranium in bismuth in power reactors is possible, The uranium is used as the reactor fuel and the bismuth is a carrier. This metal solution is pumped from the reactor to a heat exchanger and then back to the reactor. Bismuth has a low neutron capture cross section and does not attack graphite which may be used as a moderator in this type of reactor. Heavy and ordinary water may be used as primary coolants in homogeneous reactors; the fuel being a uranium salt. The water then also serves as a moderator. 64

The properties of some coolants are given in Table 13 and in Figure 17. More extensive information on the properties of liquid metals may be found in Liquid Metals Handbook, R. N. Lyon, ed. (obtainable from U. S. Government Printing Office). D. Structural Materials Consideration of the choice of reactor structural materials depends on the size and type of reactor to be built, the intended service, and the operating temperatures. Thermal reactors allow use of only those materials with low thermal neutron absorption cross sections. This restricts the number of possible structural materials which may be used in thermal reactors considerably, although fast reactors permit choice from a greater variety of materials in this respect. The economic design of reactors which are to operate at high temperatures narrows the field of structural materials since many metals which are satisfactory at low temperature (such as aluminum, for example) have melting points too low for high temperature use. Among the factors to be considered in choosing reactor materials are: mechanical quantities such as tensile strength, hardness, ductility, etc.; chemical properties such as corrosion resistance and chemical reactivity with the coolant, moderator, fuel, etc.; thermal properties like the melting point, specific heat, heat conductivity, the thermal coefficient of expansion, and solid state phase transitions; nuclear properties such as the neutron absorption and scattering cross sections (the specific gravity weights the effectiveness of these cross sections); radiation damage, which is the effect on the chemical and mechanical properties of the material on exposure to neutrons, beta and gamma rays; and there is, of course, the very important matter of cost. Table 14 shows many of these properties for some reactor materials. 65

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In regions in the reactor where the neutron flux is very high, some of the more expensive reactor materials, such as zirconium, may be used and then more reasonably priced materials employed outside the core. The use of small, high temperature reactors may provide a saving on the expensive materials. At any rate, in the overall reactor design, a compromise must be made on reactor size, operating temperature, fuel inventory, etc., and a very important factor in arriving at a sensible compromise is the properties and cost of structural materials. E, Radiation Damage Radiation damage is the term applied when the mechanical, chemical, and nuclear properties of a substance are altered by exposure to neutrons, beta and gamma rays, and fission fragments. Organic compounds suffer greatly in exposure to all these radiations. This means that hoses, hydraulic fluid, organic electric insulators, plastics "Dowtherm", etc., may not be used in regions where neutron flux or radioactivity are high. Water is readily dissociated under irradiation and provisions have to be made to re-combine the hydrogen and oxygen either through the use of a catalytic agent or direct burning. (Dissociated water is highly corrosive to metals at high temperatures.) Deuterium, either as heavy water or deuterated compound, suffers slightly through the photodissociation of the deuteron into a neutron and a proton by reactor gamma rays. Ceramics and glasses are more resistent, but most resistant of all are the metals, mainly because of the lack of covalent bonds. Where fast particles strike a metal, atoms are displaced from their original position. This leads to a change in conductivity, creep, hardness, etc. These changes may be corrected by annealing. Radiation damage has the effect of cold working the metal, 69

except that the radiation effects are nondirectional and the radiation induced hardenings is not as great as occurs in the case of cold working. A specifically nuclear effect that weakens the metaL is neutron capture reactions which may eventually result in the formation of atoms of new elements. Carried to the extreme, this eventually has the effect of weakening the metal due to the presence of impurities. F. Shielding Reactors must be shielded to protect operators and other personnel from the intense neutron and gamma ray fluxes that issue from the reactor core. In addition, shielding performs the important service of protecting organic compounds in gaskets, hoses, etc., from radiation damage. It is also important that the steam driving a turbine be shielded to prevent dissociation of the water into free oxygen and hydrogen peroxide which are highly corrosive at the temperatures of superheated steam. 1 Charged Particles The charged particles which result from radioactive decay are most easily stopped and require the least consideration in shielding calculations. Alpha particles such as occur in the radioactive decay of heavy nuclides, may be stopped in a few centimeters of air or a few thousandths of an inch in metals. Beta particles are somewhat difficult to stop but do not present a very difficult problem. The range of betas in some materials is plotted vs. energy in Figure 18. Gamma rays and neutrons are the most difficult to stop of all. The reason for this is that they are uncharged and there is For a more detailed account of the effects of radiation damage to metals and non-metals, see Nucleonics, September, 1954. 70

no electrostatic interaction with the shielding material which in the case of charged particles, provides the slowing-down mechanism. 2. Gamma Rays When the radiation from a point source is absorbed catastrophically, i.e., the particle or gamma ray is lost entirely from the flux by a single collision or event, the law which describes the decrease in intensity with distance is: r= S e2 ^o 2) 4n x2 wh3re S is the point source strength in particles per second, I is the flux in particles per square centimeter per second at a distance X from the source,'.." is the thickness of the shielding material measured along the line between the source and the point at which the flux I is to be determined, and E is the macroscopic cross section for catastrophic absorption of the radiation or particle. Z is given by N I, where CT is the cross section for the absorption due to a single atom and N is the number of such atoms:er unit volume. (Units must be chosen so that the product ZXo is dimensionless) N is given by (N ) where No -.02 x 1023 (Avogadro's number), d is the density of the absorbing medium and A is its atomic weight. From this shielding law, it is seen to be advantageous to have the product Z CX large for good shielding. If source of radiation is not a point source but is distributed in space, we must ad.d the contributions from all the radiation source material. This means we must compute: 71

FIG. /8 THE MAXIMUM RANGE OF BETA PARTICLES AS A FUNCTION OF ENERGY 100011 AIR _ ~-.I0_ _ -..- ^' AIR 0 LEAD _1 —.2.3 4.5.6.7.8:91.0 2 ~ _ I. i i /' - g ^i - - -.001.2 5.3 4.5.6 7.7B.9 -1.0.2 3 ENERGY (MEV.) From Stanford Research Institute Report No 361 "The Industrial Uses of Radioactive Fission Products"

I - ~w JjJ ~' e <xv 3) where (r ) is the source density in particles emitted per second per unit volume at the point r, % is the distance from r to the point at which I is measured, and d^-v is the differential volume element. Evaluation of the integrals involved in particular source configurations is generally difficult or impossible to do exactly, but may be approximated numerically. For gamma rays, the cross section a- for absorption depends on the energy of the gamma involved and on the shielding material. Materials with large atomic numbers are more effective as shields since C is larger. Below in Table 15 is a list of values of Z in cm for three common shielding materials. Some useful numbers which point out the relative effectiveness of shielding materials are the thicknesses required to reduce the intensity of 2 Mev gamma rays by a factor of 10o This intensity reduction is produced by 48 cm of water, 25 cm of concrete, 6.7 cm of iron, or 4.3 cm of lead. The gamma rays from a reactor can be roughly computed using Table 16. The gamma rays considered include the prompt gammas, The Science & Eng. of Nuclear Power, Vol. II, C. Goodman, ed., Addison Wesley Press, Inc., 1949, p.201. More extensive data may be found in Radiological Health Handbook, obtainable from Public Health Service, Cincinnati, Ohio, p. 124 Experimental Nuclear Physics, Vol. I, E. Segre, ed., John Wiley & Sons, Inc., New York, 1953, p. 309; Introduction to Nuclear Engineering, R. Stephenson, McGraw-Hill, 1954, Chapter 5. 73

TABLE 15 GAMMA RAY ABSORPTION COEFFICIENTS IN CM-1 FOR IRON, LEAD, AND CONCRETE 1 (cm-) Energy (Mev) Fe Concrete Pb 0.04 25 7 0.06 10 1.05 0.08 4.8 0.55 0.10 3.1 0.45 75 0.20 1.5 0.27 17.5 0.40 0.7 0 20 4.5 0.60 0.63 0.17 2.1 0.80 0.52 0.14 1.32 1.00 0.47 0o11 1.0 2.00 0.55 0.09 0.53 3.00 0.45 4.00 o.45 6.oo 0.5 8.00 0.56 TABLE 16 REACTOR GAMMA RAYS Energy Interval No. of Gammas Total Energy in (Mev) Per Fission Interval (Mev) 0-2 9.31 9.31 2-4 0.75 2.25 4-6 0.099 0.495 6-8 0. 0.0154 0.1078 8-10 0.0029 0.0241 74

fission product gammas and also fuel capture gammas. The number of gamma photons produced per second is proportional to the number of fissions per second or the power level. Figure 19 shows the relation between the fission rate and power level. 3. Neutrons The neutron flux emerging from a given reactor is proportional to the power level at which the reactor is operating. Reactor construction and geometry define how great this flux will be, and there is no simple relationship that can be stated between reactor size and the neutron leakage flux. The history of typical neutrons emerging from a reactor is of interest. A slow neutron will be captured after traversing a small thickness of shielding material, one or more capture gamma ray photons being emitted in the process. A fast neutron may be captured while it is still fast or it may be slowed down by the moderating action of the shielding material and eventually be captured as a slow neutron. A conservative estimate of the neutron shielding required may be made if the moderating property of shielding material is neglected and only fast neutron capture is considered. This is a conservative estimate since fast neutron cross sections of elements are much smaller than if taken at thermal energies. Since fast neutron capture is a catastrophic process, the same law given above for gamma rays applied except that Z now refers to neutron capture by the shielding material and is given by (N -a) where'a is the neutron absorption cross section in barns. Some values of Z are shown in Table 17. From Introduction to Nuclear Engineering by R. Stephenson, McGraw-Hill Book Co., Inc, New York, N. Y., 1954, p, 209* 75

/ / o / o, -0~ w' / i w:~ (I) / z < zIV/ o z c / 0 W', Z zr/ / w. PS / / w o (w U. _ o O,T =o 3 ^ /4" / / o! i. i_ m~ -'U./ ne _U LL / U. / /O~~~I / // / /ZO0o / / / // /.. 0 8AV*93 0 go/ ~~~~/, ~ S. VIS~~VMV93VI / ~ 0 0 / /'o /_ o / O U/ /'IH /' n'Se / / 9K - o 9/ / / / / (o o 2 o

TABLE 17 NEUTRON SHIELDING DATA Neutron Shielding Thickness to reduce Energy Material z cm Intensity by 1/10, cm. Hydrogen (in water) 0.281 8.2 1 Mev Oxygen (in water) 0.268 8.6 Lead 0.178 13 Hydrogen (in water) o.064 36 10 Mev Oxygen (in water) O.050 46 Lead 0.165 14 Tables such as this are easily constructed using neutron cross section data such as presented in "Neutron Cross Sections" AECU-2040. A nucleus which captures a neutron usually emits gamma rays which must be considered in the shield design. Ideally, a material such as concrete composed of light to medium weight nuclei to slow down neutrons should be used in conjunction with a heavy metal for gamma shielding where the heavy metal is dispersed in the concrete. Some shielding concretes made up in this manner are listed in Table 18. The neutron and gamma ray fluxes allowable outside the shielding must be less than certain maximum allowable values which are described later in this section. A problem to be considered is the heating of the shield by absorption of the neutrons and gammas. Sometimes a water cooled lead shield, close to the reactor core, is used to absorb a large fraction of the energy so produced. A concrete shield outside then reduces the *l flux down to a safe level. A description of the shield heating is given in Introduction to Nuclear Engineering by R. Stephenson, p. 217. 77

TABLE 18 HEAVY-AGGREGATE CONCRETES AND OTHER SHIELDING MATERIALS Density Cement Aggregate (gm/cm ) | |Portland Sand, gravel 2.3' Portland Barytes 5.5 Portland Iron punchings 6 Portland Barytes-colemanite 3.2 Cast Iron 7.85 g3 J Lead 11.2 Uranium 19.0 H L Thorium 11.4 4. Dosimetry The curie is that amount of radioactive material in which 3.7 x 1010 disintegrations occur per second. Knowledge of the halflife of a radioactive nucleus permits calculation of the number of curies per gram of the radioactive nuclide in question. The nomograph of Figure 20 shows how this is done. It is possible to convert the fission product yields given in Figure 7 to curies using this nomograph and the known half-lives of the fission product nuclides such as given in Table 2. Where gamma rays are emitted in the radioactive decay process, we may obtain the photon flux in (photons)/(cm -sec.) from the number of curies in a sample through use of Equation 2. Next, Figure 21 may be-used to compute the dose rate in roentgens per hour at a given distance from the source. Consideration of the amount of gamma radiation or neutron flux that the human body can tolerate without producing sickness or more permanent damage is the basis on which all reactor shields must be designed. A detailed treatise of the present status of the allowable dose of a penetrating radiation is beyond the scope of this 78

Ir tS I ( J zr S z 5.n'Z n S 3Z -J' 4 U) -J z MJ'<l - -- 10 tu: <oo' 1 r> W --, i, U. o w C5 - < W W "~ 4c go /w<SZl o j 0 43 4 Z O Pto _., o o. ) CDC WO W_0 - 0 0 WW N8 4 W I W C r W a o.. W X. W (,,,' co W I I-I'.0-.",o'0 "o\' "o o N _ 0o 0 04 000'0 w Q r,,,,'"0 - on ~ ~ ~ / -I I ""V~o ~- o J / // / / 0 PPP~ I3 / 0 - ~94: 3 r o (u M' 0 o o0 / 0, / " b I 1(', - I II' (..- I I I I / I 9.8 ~,' ~l'I~~~~~~~~~~~~~~~~~~~~~~~~

GAMMA RAY FLUX _=' ___ _ _WHIGH YIELDS ONE ROENTGEN — /~ ~ ~~ ^\ 1 P E R H O U R IN A IR co / LIo Msio'l 1/ IIl, = ^=====:::====. == = = = ==R o ==- -__===_ — 0 -00 /' - -__ =_ —— S __ x S, O ___ II11I__ - - I- I _____I_1 — 11 1___ 0 - ----. — - - - 0.01 0.1 I 10 GAMMA RAY ENERGY - MEV F/IG 21

* article; however, a brief treatment is included here. The human body can tolerate being exposed to weak radiation over an extended period of time so that the accumulated dose may be quite large although, if the same dose were administered in a short time, it might be lethal. The human body has the ability to repair itself from minor radiation damage if given time to do so. Realization of this has resulted in the development of standards defining the time average exposure tolerable and also the maximum dose taken in a short time regardless of the long time average. (See Table 19). The energy necessary to disable a tissue sample is curiously not enough to cause any appreciable heating. This suggests that the damage mechanism is much more subtle. It is believed that the mechanism by which radiation damages living tissue is the ionization of atoms in complex organic molecules in the tissue cells. There are many ways in which the ionization can take place. Gamma rays produce photoelectrons in tissue which may then cause ionization. Beta particles, protons, and other charged particles ionize material directly as they slow down. Neutrons travelling through tissue collide with protons in the organic tissue molecules to give these protons enough kinetic energy so that they ionize surrounding tissue. When a neutron is captured by nuclei in tissue, capture gamma rays are emitted which may cause further damage. The unit of exposure is the roentgen (r). This unit was originally devised to measure X-ray dose and is the amount of radiation necessary to produce 1 esu (electro-static unit) of charge in one cubic centimeter of dry air at 0~ C and 760 mm pressure. To state this definition in a manner which is independent of the * See, for example, National Bureau of Standards Handbook 59 581

temperature and pressure, this is the radiation necessary to produce 1 esu in 0.001293 grams of dry air. To produce one esu of charge in air, 83 ergs are absorbed from the radiation. The number of gamma 2 ray photons per cm -sec. necessary to yield one roentgen per sec. is dependent on the gamma ray energy. (See Figure 21). The photon flux producing one roentgen per second or 83 ergs per gram-sec. will produce 93 ergs per gram-sec. in water or soft animal tissue. The roentgen equivalent physical or rep is a newer unit of radiation dose than the roentgen and is defined as that quantity of radiation which produces 93 ergs per gram in tissue.* The rep for beta rays, protons, and neutrons corresponds to a different flux of such particles than in the case of gamma rays. It has been recommended that the rad be accepted as the unit of dose where one rad corresponds to an energy absorption of 100 ergs per gram. Another unit which is defined in terms of biological effect, rather than through energy absorption, is the roentgen equivalent man (mammal) or rem. The rem may be defined as that dose of an ionizing radiation which has the equivalent biological effect on man (or a mammal) as a dose of one rep of gamma rays. The relationship between the rem and the rep or rad is expressed by a multiplicative factor, the relative biological effectiveness (RBE). dose in rems = (dose in rads or reps) x (RBE) Often seen in the literature is the figure 83 ergs per gram in the definition of the rep. This figure is not so convenient to work with as the 93 ergs per gram figure as gamma ray flux meters which measure the ionization in an air chamber are calibrated in roentgens. The conversion to rep's is 1:1 if the 93 erg figure is used in defining the rep. 82

It makes little difference whether reps or rads are used in this relationship since they correspond to nearly the same energy absorption per gram. The RBE's of several ionizing radiations are given in Table 19 along with some other tolerance data. TAB'LE 19 RADIATION TOLERANCE LEVELS (Chalk River RBE Conference 1949) X-rays, gamma rays, beta rays 1 Thermal neutrons 5 Fast neutrons 10 Alpha particles 20 Max. tolerance dose at U. S. Atomic Energy Commission facilities 0.100 rem/day or 0.300 rem/week Background radiation to which man is exposed. Includes effects of cosmic rays, natural radioactivity of the earth, etc. 0.001 rem/day Fatal dose applied to the human body in a short time 200 to 800 r The neutron flux tolerance level is dependent on neutron energy. Figure 22 shows recent estimates. 85

Z.0 0 W ~ O U) U. (z ([z 1 0 0 0 _ I / 0 0g 2

Part III - NUCLEAR ENGINEERING 1. REACTOR TYPES AND FUELS A. Thermal Reactors A uranium-graphite pile is a graphite moderated thermal reactor which is built by stacking graphite blocks containing holes for the insertion of uranium fuel elements and control rods. The swimming pool reactor consists of an active lattice of fuel elements and control rods which is suspended in a deep water filled concrete tank. The water acts as both a moderator and a coolant. If this reactor is made slightly supercritical, the water locally expands, boils, or decomposes due to radiation damage and the resulting average lower moderator density allows a greater proportion of neutrons to escape the fuel region with the result that this type of reactor is self-regulating. The water boiler reactor is a homogeneous reactor consisting of a solution of uranyl sulfate or nitrate in light or heavy water. The solution, or soup, is contained in a spherical metal tank with provisions for cooling and control. The water in the soup is the moderator. In principle, the soup could be used as a coolant. Properly designed, a water boiler reactor is self-regulating the same way a swimming pool reactor is. Studies are being conducted at Brookhaven National Laboratories on the plausibility of building thermal reactors using uranium fuel dissolved in molten bismuth coolant. It is quite possible that thermal. reactors will never be successful as breeders with net fuel gain since the absorption of thermal neutrons by the large quantities of moderator, structural materials, and coolant nuclei tend to destroy the slim margin of breeding possibility. 85

The reactors shown in Figures 23a and 23b are thermal reactors. B. Intermediate Reactors Reactors can be built in which fission is mainly caused by neutrons in an energy range between thermal neutron energies and the higher energies which neutrons produced in the fission process have. (See Figure 6). Such reactors are called intermediate reactors. Since the fission cross sections of the fuel nuclides are lower at intermediate neutron energies than at thermal energies, a larger inventory of fuel is necessary to sustain the chain reaction in these reactors. This disadvantage is off-set by the wider variety of structural materials and coolants which may be used as compared with thermal reactors. The reason for this difference is that at intermediate and higher neutron energies, the neutron absorption cross sections of many readily available materials are low enough so that their use does not seriously poison the reactor. The slowing-down of the fission neutrons to intermediate energies is effected by the structural materials, coolant, and fuel rather than by use of an auxiliary moderating substance such as graphite or hydrogenous materials. It is possible that intermediate reactors may succeed as breeders since the neutron absorption by reactor materials is not as great as occurs in thermal reactors. Figure 23c shows an intermediate reactor system. C. Fast Reactors Few fast reactors have been built and information on the operation of these is mainly classified. Interest in fast reactors is partly due to the variety of structural materials that may be employed, as in the case of intermediate reactors, but even greater interest resulted when it was shown with the Experimental Breeder Reactor (EBR-I) at 86

- Z J 4 0 ie00 W z I4c i: c 2 e W 0 ~,.-, 11 1Z vl W - 10 W) c Ijl s.2-11 rccti II! = lb W8 8 ~ o:o 2 IL 4 10 W W1W nn3h, a Wc - \W a j 0 Ej Z4

Arco, Idaho, that more fissionable fuel could be produced than consumed in a fast reactor. This reactor uses U-255 as fuel with a core about the size of a football. Plutonium is produced in a natural uranium blanket surrounding the core. Fast reactors must be small to keep the fuel inventory low, which incidentally is larger than required by either thermal or intermediate reactors. With small reactors, there arises the problem of removing large amounts of heat from small areas. This points to the use of liquid sodium or Na-K as a primary heat transfer medium as these materials have large heat transfer coefficients. It may be possible that the attractive breeding gain of fast reactors will more than justify the inconveniences, in which case we shall likely see many power plants employing fast reactors in the future. (See Figures 23c and 23d). D. Reactor Fuel Elements and Fuels for Homogeneous Reactors Reactor fuels may be prepared using any desired mixture of fissile and non-fissile isotopes of the fuel in question, since the chemical and mechanical properties of chemical elements are independent of the isotopic constitution. Chemically, the fuel may be in the form of metals, oxides, metal solutions, aqueous solutions of fuel salts, etc., as the nuclear properties of the fuel nuclei are, for all practical purposes, uninfluenced by their chemical environment. A reactor fuel element is a piece of material containing macroscopic quantities of a fuel nuclide for use in a heterogeneous reactor. This term is not to be confused with the meaning of "element" taken in the chemical sense. A number of different designs for fuel elements have been used. In general, it is desirable to enclose the fuel in a jacket The ratio of fissile to non-fissile nuclides is dependent on the reactor design. 88

which does not allow the radioactive fission products to contaminate the surrounding parts of the reactor. The jacket must have high conductivity to allow escape of the fission heat, must be resistant to corrosion by the coolant, moderator, etc., and have a low neutron absorption cross section. Metallic uranium is often clad with aluminum for this service and is useful so long as the melting point of the aluminum is not reached in operation. Fuels for homogeneous reactors may be in the form of solutions of uranyl sulfate or uranyl nitrate in H O or DO0, solution of uranium metal in molten bismuth, etc. A high temperature reactor system under study proposes using UO particles in a bed which is fluidized by an air stream. This fuel has properties in common with both homogeneous and heterogeneous reactor fuels. 2. POWER FROM NUCLEAR REACTORS: THE NUCLEAR MANUFACTURING PLANT A. Introduction Since the U. S. Atomic Energy Commission released classified information on nuclear reactors to industrial power study groups, a variety of power plant designs have evolved, although at the time of this writing, none of the proposed plants have been built. Perhaps the most attractive idea which nuclear power plants offer is the possibility of obtaining so much energy from a given mass of fuel. Using the conservative figure of 182 Mev released in the fission of one U-235 nucleus, we readily see that the fission of one gram of U-235 releases 7.0 x 107 B.T.U. or 0.86 megawatt-days of energy. It has been estimated that a cubic block of U-235 metal eight feet on a side would provide all the world's fuel needs for a year. 89

Figures such as these permit consideration of designing ships which are to operate for long periods without re-fuelling and having the additional advantage of providing greater cargo capacity which in conventional coal and oil burning vessels is involved in the storage of fuel. Long range nuclear powered aircraft may make an appearance in the future as well and studies are presently being made in this direction. There is also the possibility of building electric power plants for remote locations in which nuclear fuels will be used with the advantage that fuel may be transported much more conveniently than is possible with conventional fuels. There is the further prospect of finding industrial uses for the intense radioactivity of the fission products. A great deal of exploratory work is now in progress with the aim of finding new large scale markets for the fission products, which may become valuable as by-products of power production. This subject is discussed further in Section 4B. Serious consideration must be given to the present rate of depletion of the world's supply of fossile fuels. A recent survey of the rate of the world's population growth, the present and estimated rate of increase of fuel use per capita, and the estimated reserves of the fossile fuels, points out that new fuel resources must be developed. Figure 24 presents the results of some of these estimates of future fuel requirements and known fuel reserves. The vast low grade reserves of the fissionable nuclides gives us hope that the high standard of living now enjoyed in the United States and some other countries will not have to be sacrificed in the next few centuries on account of power shortage. B. Power from Nuclear Reactors The energy produced in the nuclear reactor must generally be abstracted in the form of heat. This heat may be either used at low Palmer, Putnam, Energy in the Future, D. Van Nostrand and Co., New York, 1953. 90

_ _ _ _ _ __........... —------ ------— 1-1-11 11,000 Q ENERGY IN THE RESERVES OF URANIUM AND THORIUM SO FAR MAPPED AND RECOVERABLE FOR LESS THAN o100 A POUND OF METAL CONTENT AMOUNTS TO 1700 Q. BREEDING TO THE LIMIT OF ECONOMICALLY - 588 QJUSTIFIABLE CHEMICAL RERUNS WILL RECOVER SOME 575 Q. _ THE RESERVES OF ENERGY IN THE FOSSIL FUELS, RECOVERABLE 4 BY PRESENT METHODS AT NO MORE THAN TWO TIMES PRESENT REAL UNIT COSTS, AND ADJUSTED FOR LOSS-IN-SYNTHESIS, CONTAIN / ABOUT 27 Q. ---, -- -, - -- -, --. -- -- - -~3 / --- - 100 Q S mt I I L 1 I I I I 1 r I I r~ c I I I 1 1 A or 5 u Lif ~~0~ ~. X__ _.....- 4 I0 x F —~ — 1 10 0 "l - - - - - ~~HYPOTHETICAL WORLD POPULATION IN A.D. 2050 8 BILLION POPULATION l — ---- 5 BILLION POPULATION.,,_ ___ I I______________,_______I____________ —-— I I Q 1850 1900 1950 2000 2050 YEAR FIG. 24 Estimates of maximum cumulative World demand for energy input, A.D. 1947 to 2050, assuming two different maximum plausible populations, three different maximum plausible rates of growth in the per capita demand for energy output measured at the point of end use, and one minimum plausible trend in the weighted average world efficiency curve. The value of 9 Q at A.D. 1860 represents the approximate cumulative input to the World energy system between A.D. 1 and the earliest date for which there exist reasonably adequate estimates of annual fuel consumption. The demands are compared with the 27 Q recoverable from fossil fuels and the 575 Q recoverable from nuclear fuels, at less than 2 times 1950 costs. From "Energy in the Future" by Palmer Putnam, D. VanNostrand Company, Inc., New York 1953.

temperatures for building heat, or it may be used at high temperatures to produce steam which may be fed through a turbine to produce electrical power. A third possible means of using energy from a nuclear reactor is as high temperature heat for the initiation of chemical reactions or the cracking of petroleum, or other high temperature requirements such as metallurgical operations. The choice of which of these modes of heat utilization is to be adopted has a profound influence upon the design of the reactor. If heat is merely to be abstracted in order to prevent destructively high temperatures in the reactor, then the heat may be discarded at low temperature in cooling water or air, or may be used for some purpose such as building heating. A reactor operated in this manner has less stringent requirements upon the physical properties of the structural and coolant materials than does one operated at a higher temperature. However, if one is to employ the energy for the generation of power, then a relatively high temperature must be produced in the reactor. Liquid metals may be used as coolants as shown in Figure 23c and 23d, or water at high temperature and pressure may be used as shown in Figure 23a. In either of these two alternatives, the coolant channels within the reactor must be designed to withstand the corresponding chemical and physical conditions. In addition, care must be taken that the fuel within the reactor may transfer heat to the coolant through sufficiently short distances so that destructively high temperatures are not reached within the fuel itself. Coolant materials passing through the reactor core will in general absorb some neutrons, and may become radioactive. Consequently, it has been proposed that the coolant not be charged directly to the turbine or to other heat utilization devices, but rather should be caused to exchange its heat with a secondary coolant loop. A secondary coolant, 92

although exposed to the radioactivity of the primary coolant, would not become radioactive, and consequently could be used in machinery without fear of radioactive contamination. Such a secondary heat exchange system is shown in Figure 23c. Liquid metals have been proposed as a primary reactor coolant because they may attain relatively high temperatures without correspondingly high vapor pressures. These liquid metals, such as sodium, may then exchange their heat with a secondary coolant which would be suitable for operation of a turbine. It appears possible that the secondary coolant might also be a chemical or petroleum stock which is to be subjected to high temperature in order to initiate cracking or other chemical change. Inasmuch as radiation often initiates chemical reactions, it may be possible to obtain products with properties resulting from a combination of the thermal and radiation effects. Such properties may not be attainable in present thermally initiated reactions, and therefore offer promise of investigation. C. The Nuclear Manufacturing Plant Nuclear Reactors, together with their attendant auxiliaries, can be considered from one point of view as manufacturing plants. There are three chief products of such a plant, namely: 1. Energy which appears initially in the form of radiation and is degraded to heat and may ultimately be converted to electric power. 2. The fissionable materials, uranium23 or plutonium239, which are produced by the absorption of neutrons in source material. 3. Gross fission products. One kind of raw material for these plants are source materials such as uranium238 and thiumum232 which, upon the absorption of neutrons, become converted to fissionable materials. Another kind of raw materials 93

are the fissionable materials themselves which may be regarded as the fuels for the manufacturing plant. If one is to consider a nuclear reactor as a part of a manufacturing plant, it is necessary to consider sources of raw materials, the operation of the reactor and its auxiliaries, and markets for the products of the manufacturing plant. The sequence of interdependent operations upon which the nuclear manufacturing would be based are the following: 1. The preparation of feed materials, which are the source and fissionable isotopes required for the operation of a reactor. 2. The enrichment of naturally occurring ruanium by means of gaseous diffusion to U-235 concentrations sufficiently great that it may serve as a fuel for a reactor. 3. The formulation of source and fissionable materials into physical and chemical states suitable for use as nuclear fuels. 4. The operation of a nuclear reactor. 5. The conversion of the energy from the fission reaction to electricity or to heat for other purposes. 6. The chemical or physical separation of irradiated fuel to permit the recycling of unused source and fissionable material for further use as fuels. 7. The processing of separated fission products with the production of packaged gross fission products or the isolation and packaging of selected individual products. In Figure 25, a hypothetical scheme is shown which would permit conducting the operations just described. According to this scheme, natural uranium containing 0.71% of the fissionable isotope uranium235 is the raw material input to an enrichment plant. Partial enrichment of 94

FIG. 25 SPENT FUEL GRAMS/DAY % REACTOR FUEL U-235 23 0.11 GRAMS/DAY % U-238 20,250 208 1.0 Pu 127 0.61 20,592 F.P. 400 1.92 20,800 20,800 NATURAL U GRAMS/DAY % U-235 219 0.71 / \ REACTOR U-238 30,581 30,800 / ISOTOPE \ HEAT | 400 MW ~ SEP'N \. PLANT POWER PLANT 25% EFF'Y DEPLETED U GRAMS/DAY % U-2 35 I 0o.11 U-2358 9989 011 POWER 1 100 MW U-258 9,989 10,000 ENRICHED URANIUM REACTOR SYSTEM WITH ENRICHMENT PLANT PRODUCTS: FIGURES 25-28 FROM: POWER ENRICHED URANIUM M. BENEDICT, IND. a ENG. PLUTONI UM DEPLETED URANIUM CHEMISTRY, NOV. 1953. RADIOCHEMICALS

the fuel would occur, resulting in the production of two streams from the isotope separation plant. The enriched stream contains 1% of uranium235 and the depleted streams 0.11%. The enrichment of the fissionable isotope from 0.71% to 1% in concentration imparts to the enriched fuel a sufficient reactivity in excess of that occurring in natural uranium to permit certain additional latitude in reactor design and operation. The products of the reactor described in this system will be plutonium, fission products and power. The production of power from heat is assumed to be straightforward, and it is assumed in this operation that the steam turbine and generating equipment have an overall efficiency of 25% based upon the incoming heat. The spent fuel discharged from the reactor contains a residual 0.11% of uranium and 0.61% of plutonium. The total of these is 0.72%, which is not as great as the 1% of U235 which entered the reactor. Consequently this type of reactor does not produce as much fissionable plutonium as it burns uranium255. The operation of this type of system depends upon the availability of an isotopic separation plant, since 255 the uranium5 which is produced is consumed without the production of corresponding quantities of plutonium. The plutonium itself might be used as a fuel for recycling through the reactor if it were produced in 255 quantities sufficient to replace the uranium235 originally charged. The depleted uranium which is produced by the isotope separation plant is a by-product of this system. An alternative scheme, calling for the operation of a reactor with natural uranium raw material, is given in Figure 26. In this scheme, natural uranium is mixed with plutonium from previous reactor operations. Plutonium might have been produced previously in some reactor employing a cycle using no recycle plutonium, such as that described in Figure 25. Once a supply of plutonium is available, it can be used to enrich the 96

PLUTONIUM |~PLUTONIUM SPENT FUEL 90 GRAMS/DAY GRAMS/DAY % U-235 5 0.04 U-238 13,705 PU 90 0.64 F.P. 400 2.81' 14,200 REACTOR FUEL CHEMICAL REACTOR _ SEPARATION PLANT HEAT 400 MW REACTOR WASTE POWER PLANT 25% EFF'Y NATURAL U POWER I100 MW NATURAL U GRAMS/DAY %/ U-235 100 0.71 U-238 14,010 FIG. 26 14,110 PLUTONIUM RECYCLE REACTOR SYSTEM USING NATURAL URANIUM FUEL PRODUCTS: POWER PLUTONIUM RADIO CHEMICALS

incoming natural uranium with fissionable plutonium instead of adopting the expedient described in Figure 25 of conducting an isotopic separation upon a natural uranium in order to increase the content of fissionable uranium 5. During operation of the reactor of Figure 26, the 235 plutonium and uranium which are charged, undergo fission, releasing 2358 neutrons which are absorbed in uranium8 and thereby, producing more plutonium. Some of the plutonium thus produced fissions in turn and produces more power and more plutonium. This sequence of events may be continued until the presence of fission product poisons necessitates the removal of the fuel because of declining reactivity. At the conclusion of this reactor cycle, there remain a total of 95 grams of fissionable material compared with the 190 grams which were charged to the reactor. Consequently, a system of this kind requires the continual charging of fresh natural uranium, and hence, causes a gradual depletion of the stock of fissionable 235. However, if the reactor described here were to operate as shown, no net depletion of the original charge of plutonium would occur. The products from such a system would be power, plutonium to sustain reactor operations, and radiochemicals. Figure 27 describes a scheme for self-sustaining nuclear manufacturing system. This system produces slightly more fissionable material than it consumes, and differs in this respect from the systems described in Figure 25 and 26. Consequently, the only raw material required is uranium2 8. The uranium238 could be taken in the form of depleted uranium from the operations of an isotope separation plant such as that described in Figure 25, or from the discharged fuel from a reactor such as that of Figure 26. The scheme shown in Figure 27 is a U38 - Pu fast breeder reactor. This scheme is more complicated than either of the 98

PLUTONIUM 401.4 GRAMS/DAY U-238. 39703 GRAMS/DAY U-238 AU-238 REACTOR 1% Pu BLANKET 400 40,103 GRAM BLANKET SEP. PLANT GRAMS DAY NEUTRONS 1.4 GRAMS/DAY FISSION PRODUCTS 400 GRAMS/DAY REACTOR CORE PU 2%FP 20,000 GRAMS CORE — SEP. PLANT DAY I___! __s___.__' " HEAT | 400 MW POWER PLANT 25% EFF'Y POWER 100 MW FIG. 27 FAST PLUTONIUM U-238 BREEDER REACTOR SYSTEM PRODUCTS: POWER PLUTONIUM RADIO C HE MICA LS

previous reactor systems described. Most of the fission occurs in the reactor core. Neutrons escaping from the surface of the core are absorbed in the surrounding blanket of uranium238 instead of being allowed to escape as in the previously described systems. The capture of these neutrons accounts in large part for the fact that more fissionable material is produced than is consumed in this type of system. The course 258 of operations is that uranium is brought into the system and mixed with recycled uranium238 from the blanket separation plant. This mixture is then fed through the blanket, whereupon absorbing neutrons, some of the U238 is converted to plutonium. This plutonium is removed upon subsequent separation. Some of this plutonium may be mixed with plutonium separated from previous core operations and fed to the core as fuel. After undergoing the reaction of fission in the core, the plutonium fuel is again subjected to the separations process to remove the fission products, and is recycled for more burn-up. Again it is considered that the operation of the power plant is of a conventional nature and is not further described here. The raw material upon which this type of plant operates is naturally occurring uranium or depleted uranium 238, and the products are power, plutonium and radiochemicals. Because construction of this type of reactor system is more complex than those of Figures 25 and 26, the capital and operating costs may be greater. This fact may offset to some extent the economic attractiveness of producing more fissionable material than is consumed. In Figure 28, an alternative reactor scheme is shown which is known as a U233 Th thermal breeder reactor system. The raw material to the system is throium232. This is mixed with a certain amount of thorium 232 containing uranium233 produced in previous reactor operations, and is charged to the reactor. During reactor operations the 100

SPENT FUEL GRAMS/DAY F.P. 400 REACTOR FEED 233 208 GRAMS/DAY TH 209 20,800 U-233 208 (1%) TH 20,592 20,800' REACT2 XOR | CHEMICAL ----- REACTOR — SEPARATI ON THORIUM 400 GRAMS/DAY FISSION PRODUCTS 400 GRAMS/DAY HEAT 400 MW POWER PLANT 25% EFF'Y FIG. 28 THERMAL U-233 THORIUM BREEDER REACTOR SYSTEM POWER 100 MW PRODUCTS: POWER U-233 RADIOCHEMICALS

reaction of fission occurs and excess neutrons are absorbed in the thorium232 rather than in uranium238 as in the previously described reactor systems. During startup operations, it will probably be necessary to use uranium235 as a fissioning fuel in order to initiate the cycle. After the reaction of fission and absorption of neutrons has occurred, the fuel is sent to the chemical separation plant, where fission products are removed and stored or sold. The decontaminated fuel contains uranium233, as described in Part II, Sec. 2E. The uranium23 and thorium232 output from this separation plant can then be recycled to the reactor for further conversion and power production. If uranium233 were produced in excess of recycle requirements the surplus would be a product of operations. The raw material required for this system is 252 25 thorium, which occurs in nature. The products are power, uranium233 and radiochemicals. It is not known for certain however, that the thermal Th-U breeding cycle is self-maintaining without the occasional addition of small amounts of U235 The choice of the system to be employed is governed as in the case of other types of manufacturing plants by many considerations not necessarily related to the technical feasibility of the operation. Among these considerations are availability and cost of raw materials, such as natural uranium and natural thorium, uranium isotopically enriched in 235 uranium25, plutonium, structural materials, coolants, moderators; the cost of operating the separation plant and many other factors which can be determined only through a detailed technical and economic analysis. It is thought that the development of breeder reactors will be a necessary part of a nuclear manufacturing economy. This will be the case if full advantage is to be taken of our natural resources of uranium. Since most of natural uranium is in the form of uranium23, the use of 102

breeder reactors should produce supplies of fissionable material in excess of those used. Such a circumstance would justify the operation of less efficient fissioning or converter reactors for power production in applications where a less complicated type of reactor installation is desirable. 3. FUEL SEPARATION A. Introduction As operation of the reactor continues with a given fuel charge, the concentration of fission product materials increases, and consequently neutron losses due to absorption by the fission products increases. Ultimately, a condition is reached in which the original fuel has produced enough fission products so that the nuclear chain reaction can no longer be sustained and the initial charge of fuel must be removed and replaced by purified fuel. Another type of behavior exhibited by metallic nuclear fuels undergoing irradiation is that of radiation damage. As a consequence of operation at elevated temperatures, and as a consequence of the production of the fission products within the crystal lattices of the nuclear fuels, the fuels themselves undergo structural changes which are deleterious to their mechanical and thermal functioning. Consequently, both radiation damage and fission product build-up constitute reasons why nuclear fuel must eventually be removed from a reactor and replaced by purified and reconstituted fuel. Fuels which are no longer suitable for reactor operation might be discarded in some suitable manner except for two reasons. The first of these reasons is that only a fraction of the fissionable material originally charged to the reactor will be consumed before radiation damage and fission product build-up necessitate removal from the reactor. Consequently, if the fuel were discarded, much valuable fissionable 103

material would also be discarded along with the waste products. The second reason why the fuel cannot be discarded is that in reactors which contain 238 258 any uranium2, the capture of neutrons by the uranium238 results in the production of plutonium which is a reactor fuel too valuable to discard. Consequently, it is found that irradiated reactor fuels must be treated in some manner so that the fissionable uranium and plutonium may be recovered from these fuels for re-use as reactor fuel. B. Aqueous Methods In the following discussion, various means will be described of accomplishing the treatment of irradiated nuclear fuels in such a manner that fissionable materials are recovered for re-use. The first method of fuel separation which will be described is aqueous chemical processing. Aqueous chemical processes are applied to the separation of uranium, plutonium, and fission products, when the component parts of the fuel must be separated in high degrees of purity. For heterogeneous reactor fuels in which the fuels are discreet bars of metal surrounded by a moderator and coolant, it is necessary to dissolve the solid fuel elements in acid. The dissolution is conducted in such a manner that the resultant solution has properties which will permit effective chemical separations. Since ordinary or light water is used in aqueous chemical separations, an ideal application of the aqueous method is the separation of fuels in a light water homogeneous reactor. If heavy water is used in a homogeneous reactor, it is necessary to remove the heavy water from the dissolved salts of fissionable material, since the heavy water is so expensive, and to redissolve the dehydrated fuel and fission product salts in light water for subsequent chemical processing. 104

The separations of uranium, plutonium, and fission products may be accomplished by one of three aqueous methods or combinations thereof. These methods are: 1 Precipitation 2. Solvent extraction 3. Ion exchange Earlier methods of aqueous chemical processing employed a precipitation technique. The use of bismuth phosphate permitted the selective separation of plutonium from a mixture of uranium and fission products. This permitted plutonium isolation in pure form, but did not permit the separation of uranium from highly radioactive fission products. A typical solvent extraction process is illustrated in Figure 29. The uranium and plutonium contained in the dissolved reactor fuels are oxidized to the hexavalent state prior to solvent extraction. In this state of oxidation, these materials may be extracted from an aqueous solution by means of solvents such as: diethyl ether, tributyl phosphate dissolved in kerosene or in carbon tetrachloride, and hexone. The various steps required in the solvent extraction process are: 1. The extraction of uranium and plutonium from an aqueous solution of fission products by means of a solvent. 2. The extraction of the plutonium salt from the uranium salt. 3. The stripping of the uranium salt from the solvent. 4. The extraction of the separated plutonium salt from small quantities of fission products. 5. The conversion of the aqueous solution of uranium salt to the form of metallic uranium and the subsequent refabrication of fuel elements from the uranium. 105

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Methods of aqueous chemical separation employing ion exchange techniques offer promise in certain uranium separations, plutonium purification, and in the isolation of specific radiochemicals from the solution of fission products. C. Fluoride Volatility Methods An alternative to the aqueous chemical processing technique for the separation of uranium, plutonium, and radiochemicals, is that of the fluoride volatility technique. The basis of the fluoride volatility technique is that uranium hexafluoride is a volatile material which condenses at conditions not far removed from ordinary temperatures and pressures. There would seem to be the possibility, therefore, of forming uranium hexafluoride from the uranium present in irradiated nuclear fuels and then distilling this uranium away from the plutonium and fission products in the form of the uranium hexafluoride. The residual materials from the distillation may be treated further as required, and the uranium hexafluoride is either reduced to the form of metallic uranium for recycle to a reactor, or the uranium hexafluoride may be charged directly to a gase — ous diffusion plant for re-enrichment of the uranium235 content. A block flow diagram of a representative fluoride volatility process is shown in Figure 30. D. Pyrometallurgical Processing A third method for the treatment of irradiated nuclear fuels to render these fuels suitable for re-use in a nuclear reactor is that of pyrometallurgical treatment. Pyrometallurgical methods have the common feature of preserving the chemical state of the reactor fuel during processing. Consequently, a minimum of chemical conversions are required in this method of fuel treatment. Alternative pyrometallurgical processes are: 107

BROMINE TRIFLUORIDE IRRADIATED FUEL MAKE UP | REACTION VOLATILE FI BROMINE PRODUCT FLORDS'"'4 - DISTILLATION PRODUCT FLUO TRIFLUORIDE URANIUM HEXAFLUORIDE PLUTONIUM' AND FISSION PRODUCT FLUORIDES AQUEOUS CHEMICAL PROCESSING PLUTO NIUM FISSION PRODUCTS FIG. 30 BLOCK FLOW DIAGRAM OF A FLUORIDE VOLATILITY PROCESS

1. Melting and re-solidification of the fuel. 2. Extraction of the molten fuel by means of metals or fused salts. 3. The distillation and condensation of fuel metals at high temperatures. 4. Zone melting. Pyrometallurgical processes are quite straightforward in concept. These processes are dependent, however, upon materials capable of withstanding high temperatures. With present knowledge of high temperature materials, melting and extraction have attractive features. Melting alone has application to fast reactors where rigid specifications of purity in the fuels are not essential. Extraction at high temperatures is a technique whereby separations between uranium, plutonium in the fission products, as well as structural materials, may be achieved. A flow diagram of a representative pyrometallurgical process is illustrated in Figure 51. 4. USE AND STORAGE OF THE FISSION PRODUCTS A. Introduction Economical nuclear power may be contingent on the successful development of large-scale industrial uses for the extensive quantities of fission products produced in nuclear chain reactors. Thus far, the fission products have constituted a high-cost liability. Their presence in reactor fuel elements requires expensive chemical separations plants to provide purified nuclear fuel suitable for re-use. in reactors. Present as wastes after chemical separation, the fission products incur an 109

IRRADIATED FUEL MOLTEN SLAGGING OR MOLTEN URANIUM' EXTRACTION FLUX SLAG CONTAINING PLUTONIUM AND FISSION PRODUCTS DISTILLATION AUXILIARY OR EXTRACTION FLUX PLUTONIUM FISSION PRODUCTS FIG. 31 BLOCK FLOW DIAGRAM OF A PYROMETALLURGICAL PROCESS

even greater expense in handling and storage. Economic studies indicate that handling and storage costs may approach 60 per cent of the total costs of chemical separations. The consumers of reactor power will bear the expense of these operations, unless methods are devised to defray the cost by industrial utilization of the fission products. Research groups from industry and the universities are presently striving striving to resolve the fission-product problem by investigating potential uses for the fission products. Many of their findings have previously been put to profitable use, and some show definite promise for future applications; others are highly speculative. The final solution to this challenging problem will determine the degree to which power from nuclear reactors will compete in a free market. B. Storage of Fission Products These materials must be maintained in biologically safe locations. If the fission products are to be regarded as wastes, one may consider dispersing them in the atmosphere or in the ocean, casting them in concrete or clay and burying them in special plots or at sea, storing them in tanks as aqueous solutions, or in the form of solids, etc. Dispersal appears to be unsafe for the very huge quantities produced by a nuclear power industry. Consequently, there is strong incentive to develop economical methods of containment. The most economical storage method would probably be that in which fission products are contained in the smallest volumes in order to minimize the costs of surrounding shielding. In the case of the long lived fission product nuclides such as 90 137 Sr and Cs continued addition of a stockpile at a constant rate will eventually yield and equilibrium value of the radioactivity (in curies) of the nuclide in question. This equilibrium value may be calculated 111

through: (curies at equilibriurm)= 8.45 x 10 3 (power level in watts)(fission yield, %) where the power level refers to the sum of the gross power levels of the reactors adding the fission product to the stockpile and the fission yield is given by curves such as Figure 7. Thus, a reactor operating at 100 megawatts eventually builds up a stockpile of 4.5 megacuries of Sr. The time required to accumulate this much Sr90 will be two to three times 90 the half-life of Sr or 40 to 60 years. C. Use of Fission Products Much heat is evolved as fission products decay and so megacurie amounts of fission products are useful as heat sources as may be seen in Figure 20. A great variety of other uses for fission products are being investigated. Figure 32 shows a list of possible uses, many of which have been investigated. Such a listing is of temporary validity, since new uses are rapidly being discovered. 112

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5. CONVERSION FACTORS To Convert From To Multiply By Ergs British Thermal Units 9.481 x O11 Ergs Gram Calories 2.389 x 10"8 -7 Ergs Joules 10 -10 Ergs per second Kilowatts 10 -6 Million Electron Volts Ergs 1.60 x 10 Million Electron Volts Killowatt-hrs 4.44 x 10 Million Electron Volts British Thermal Units 1.51 x 1016 Mev Megawatt-days 1.85 x 102 105 curies at 0.7 mev. per disintegration Kilowatts 0.41 105 curies at 0.7 mev. per disintegration Btu per hour 1400 105 curies at 0.7 mev. Lbs. steam per hr. at 1,000 per disintegration Btu per lb. 1.4 Curie Disintegrations per second 3.7 x 1010 Rad Ergs per gram 100 Roentgen Rad 0,83 114

INDEX Albedo, 38 Ion exchange, 105 Alpha particle, 6 Isobars, 2 hazards due to, 8 Isomeric transitions, 11 Atom, 1-2 Isotopes, 1-2 Atomic Bomb, 33 Atomic mass units (amu), 6 Atomic nucleus, 1-2 Jacket, 89 Atomic number (Z), 1, 8, 22 Avogrado's number, 71 Liquid hydrogen, fractional distillation of, 61 Barn, 28 Beta processes, 8 decay, 8, 22 Mass defect, 2 Beta rays, 8 measurement of, 4 Mass spectrograph, (footnote) 4 use in isotope separation process, 51 Coolants, 63-64, 85, 92-93 MEV, (footnote) 4 Curie, 78 Millibarn, 28 Moderators, 35, 37, 58-63 Multiplication factor, 34, 54 Deuterium, 58 Deuteron, 19 Dosimetry, 78-83 Neutrino, 9, 23, 26 Neutron capture, 75, 81 Neutron cross-section, 28 Electromagnetic pump, 64 1/v law, 28 Electron capture, 8 Neutron economy, 54 Exogeric, 26 Neutron flux, 75 Experimental Breeder Reactor, 52, 86 Neutrons, 57-58 delayed, 22, 57 energies, 58 Fission, see Nuclear fission prompt, 57 Fission products, 109-112 slowing down of, 58 Fossil fuels, 90 thermal, 28-29, 37 Fusion, 6, 26 Neutron temperatures, 28, 56 Nuclear cross-section, 27-28 macroscopic, 28, 71 Gamma rays, 11, 22 microscopic, 28 catastrophic absorption of, Nuclear fission, 19-26 cross-section of, 73 activation energy, 19 from fission process, 73-74 chain reaction controlled, 33 hazards due to, 8, 78 chain reactionsself-sustaining, 33 delayed neutron emitters, 22 energy available from, 6, 23 Hazards, 8, 54, 78 modes of fission, 20 Heavy water, 58, 62, 104 photo-fission, 19 115

Nuclear fuels, 38-54 Nuclear reactors (continued) carnotite, 38 types of (continued): fuel breeding, 38, 51 converter, 52 fuel enrichment, 49 -- fast, 37, 57, 86 fuel separation, 103-109 heterogeneous, 38 by aqueous methods, 104-107 homogeneous, 38 by fluoride volatility methods, 107 intermediate, 86 by pyrometallurgical processing, pile, 33 107-109 swimming pool, 85 metallic uranium, 41 thermal, 37, 57, 85, pitchblende, 38 uranium graphite pile, 85 plutonium, 46 water boiler, 85 thorium, 45 Nucleons, 4 Nuclear isomers, 11 Nucleus, 1-2 Nuclear power plant designs, 89-103 cross-sectional area of, 44 Nuclear reaction, 17-18 radioactive, 7 compound nucleus theory, 18 half-life of, 7, 78 energy abstracted from, 90-93 mean life of, 7 fission, 19-26 rate of decay, 7 fusion, 6, 26 stability against decay, 9 representation by classification, 17 alpha decay, 7 representation by equation, 17 beta decay, 9 thermonuclear, 26 uranium, radius of, 44 Nuclear reactors, 33 Nuclides, 1-2 blanket, 38 alpha particle emitter, half lives, 7 control of, 54-57 beta radioactive, half lives, 9 coolants, 63-64, 85, 92-93 decay chains, 11 critical, 34 fissile, 26 critical assembly, 35 notation, 2 critical mass, 34-35 radioactive, 2 critical size, 34 stable, 2, fuel element, 88 as manufacturing plants, 93-103 moderators, 35, 37, 58-63 Plutonium, 46 beryllium oxide, 63 Plutonium-239, production of, 54 graphite, 61 Positron, 8 heavy water, 58 Precipitation, 105 ordinary water, 61 Pyrometallurgical processing, 107-109 non-leakage probability, 34 power from, 89-93 prompt critical, 57 Radiation, biological effect of, 81 radiation damage, 65, 69 tolerance of human body to, 78-83 reactor poison, 37, 56 Rad, 82 removing heat from, 63 Radiation damage,69-103 reflector, 38 Reactivity, 56 shielding, 70-84 Relative biological effectiveness, 82 structural materials, 65 Roentgen, 81 sub-critical, 34 Roentgen equivalent man (mammal) (rem), super-critical, 34 82 types of: Roentgen equivalent physical (rep), 82 atomic bomb, 33 breeder, 52, 102, 116

Shielding, 70-85 Thermonuclear reactions, 26 charged particles, 70 Thorium metal, 45 gamma rays, 71 Transuranium elements, 46 neutrons, 75 thickness of, 73 Shielding law, 71 Uranium-235, Shielding materials, heating of, 77 electromagnetic separation of, 51 Solvent extraction, 105 gaseous diffusion, 49 Spectrum, electron and positron energies, 9 Z (atomic number), 1, 8, 22 117

BIBLIOGRAPHY Asaro, F,, and Perlman, I., "Table of Alpha-Disintegration Energies," Reviews of Modern Physics, October, 1954. Bain, George W., "Geology of the Fissionable Materials," Economic Geology, 45, 274, (1950). Glasstone, S.,, Source-Book on Atomic Energy, New York, D. Van Nostrand Company, Inc., 1950. Goodman, C., ed,, The Science and Engineering of Nuclear Power, Vol. II, Wesley Press, Inc,, 1949. Handbook 59, National Bureau of Standards, Katz, J. J,, and Rabinowitch, E., The Chemistry of Uranium, National Nuclear Energy Series, New York, McGraw-Hill Book Co., Inc., 1951. King, R. W., "Table of Total Beta-Disintegration Energies," Reviews of Modern Physics, October, 1954. Lyon, R. N., Liquid Metals Handbook, U. S. Government Printing Office. "Neutron Cross-Sections," AECU-2040. "Nuclear Data," National Bureau of Standards Circular 499. Nucleonics, September, 1954. Putnam, P., Energy in the Future, New York, D. Van Nostrand and Co., 1953. Radiological Health Handbook, Public Health Service, Cincinnati, Ohio. Rennie, C. A., Chemical Engineering Progress Symposium Series, Vol. 50, No. 12, 1954, American Institute of Chemical Engineers, p. 222. Saline, V. L., and Brown, K. B., "Studies of Recovery Processes for Western Uranium-Bearing Ores," AECD-3241, October, 1949. Seaborg, G. T., and Katz, J. J., The Actinide Elements, National Nuclear Energy Series, New York, McGraw-Hill Book Co., Inc., 1954. Serge, E., ed., Experimental Nuclear Physics, Vol. I., New York, John Wiley & Sons, Inc., 1953. Stephenson, R., Introduction to Nuclear Engineering, New York, McGraw-Hill Book Company, Inc., 1954. 118