Replacement Zircaloy for Silicon Carbide as Fuel Cladding Material in PWR – A Neutronic Evaluation

Abstract

Silicon carbide (SiC) has a potential to replacement zircaloy as fuel cladding material due to its high temperature tolerance, chemical stability and low neutron affinity. These characteristics may improve the economic and safety of nuclear reactors. Previous work has examined the possible use of SiC as a fuel cladding material in a PWR (Pressurized Water Reactor) environment. However, the advantage thermo mechanical and neutronic analysis replacement zircaloy cladding is not clear. Literature reviews has been done to predict the thermo mechanical behavior of SiC in high temperatures. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This codes system is widely accepted and used worldwide for safety analysis and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR. It was used the CSAS6 and TRITON modules. The goals are to evaluate the behavior of the infinite multiplication factor (k inf ) and neutron flux using SiC as a fuel cladding material.