Chromium-Coated Zirconium Cladding Neutronics Impact for APR-1400 Reactor Core

Abstract

The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31.