Risk Assessment and Safety Analysis of Fuel Rods in Pressurized-Water Reactors (PWRs)

Abstract

In pressurized-water reactors (PWRs), the mechanical integrity of claddings should be ensured under various loading conditions to avoid exposure of hazardous nuclear fission products to the external environment. In light of this, several problems associated with the possible failure of the cladding were studied as follows. Cracking and spalling of the oxide layer developed under normal operation on Zircaloy-4 cladding, associated with the localized development of hydride in the cladding, were studied. Through finite-element simulations, it was found that the tensile hoop stresses develop in the outer region of the oxide after a critical thickness. The conditions for the propagation of radial cracks perpendicular to the interface, channeling along the axial direction of the cladding, and the possible subsequent spalling of the oxide was then determined, using linear-elastic fracture mechanics (LEFM). Crack channeling could be classified into three different cases, depending on the thickness of oxide layer. Finally, the spalling analysis indicated that spalling leaves a thin layer of the oxide adhered to the cladding. An analysis was conducted for the cracking of Cr-coated Zircaloy claddings, proposed for use in accident-tolerant fuels (ATF) to reduce hydride embrittlement of bare Zircaloy claddings. Through finite-element calculations, it was found that uniform tensile hoop stresses within the Cr-coating layer develop by pellet-cladding interaction (PCI). The energy-release rate for crack channeling was then compared with the toughness of Cr to develop a design map for the critical coating thickness below which fracture will not occur at different levels of ramped power. In addition, a design protocol was proposed, which allows a simplified fracture analysis of Cr coating using existing PCI codes. This provides a design option without the need for constructing a full finite-element modelling of the cladding and coating. The PCI under various power histories was studied, focusing on the evolution of local hoop stresses developed at the inner surface of cladding in the vicinity of fuel cracks. If PCI is initiated either before power ramping or during the transient period of a power ramps, the local hoop stress becomes tensile, which may lead to PCI failure. On the other hand, if PCI occurs during the normal operation or steady-state portions of a power ramp, the local hoop stress becomes compressive, which suppresses possible PCI failure. These results indicate that the power history has a significant effect on possible PCI failure, which can give an insight into developing safety criteria for operation of reactors. As part of the of analysis of PCI, the fundamental contact mechanics associated with the local stress field near the corner of contact was studied. The influence of material dissimilarity on the traction fields at the contact corners between an elastic right-angle wedge and an elastic half-plane was studied through an asymptotic analysis. It was found that possible material combinations could be classified into two distinct categories. Through asymptotic analysis, a critical friction coefficient was found at which a discontinuous change from singular to bounded local fields occurs at the leading edge. Through the finite-element simulation, it was found that the contact region shrinks to zero at the critical friction coefficient, implying the existence of a concentrated force, while the singular field associated with a point contact is retained with a further increase in the friction coefficient.