The Effects of In-Reactor Irradiation on the Mechanical Properties of Zirconium Alloys in the Vicinity of the Metal–Oxide Interface

During aqueous corrosion of zirconium-based fuel claddings, the oxide grown on the outer surface protects the material from further oxidation and hydrogen uptake. Therefore, the mechanical properties and resistance to fracture of the oxide layer and the multilayer structure formed underneath the oxide, in the vicinity of the metal–oxide interface, can provide useful information about the behavior of the material. This paper builds on previous work performed on Zircaloy-2, irradiated in a boiling water reactor at two different levels: three and nine cycles. In this study, the evolution of mechanical properties of Zircaloy-2 with a different residence time (i.e., six cycles) is reported and compared to an autoclaved sample and to the previous results. Furthermore, a low-tin Zircaloy-4 irradiated in a pressurized water reactor is examined along with an autoclaved sample of the same cladding to be compared with results from the Zircaloy-2 material. Thus, not only is the same material irradiated analyzed at different cycles and compared with its archive oxidized reference, but two different materials are compared. The impact of irradiation as well as that of the type of reactor and material are discussed.