Corrosion of FeCrAl alloys used as fuel cladding in nuclear reactors

Abstract The safety of nuclear power plants has put forward new requirements for the corrosion resistance of fuel cladding with the advancement of industrialization. FeCrAl alloys are expected to be a substitute for Zircaloy. This work summarizes the corrosion of Zircaloy and presents a comprehensive review on the corrosion behavior of FeCrAl alloys including oxidation kinetics, corrosion processes and influencing factors in nuclear reactors. Corrosion performance of Zircaloy in nuclear water chemistry is definite while influencing mechanisms of relevant parameters (e.g. coolant pH) are still unclear. A small change of LiOH concentration may cause contrary effect on corrosion behavior. Different from Zircaloy cladding, FeCrAl cladding shows various kinetics in corrosive environments. A single, duplex or triple oxide layer is all possibly observed. Specific corrosion behavior depends on the properties of water chemistry and chemical reactions of main alloying elements. Developing challenges and prospects of FeCrAl alloys are concluded to expand its commercial scale in nuclear fuel cladding. For successful commercial application of FeCrAl cladding, a systematic evaluation considering corrosion resistance and other properties (e.g., mechanical strength, thermal hydraulic characteristics and neutron economy) is required.