Study of the behavior before and after ion irradiation of chromium coated zirconium alloy for use as an innovative nuclear fuel cladding in LWRs

In Light Water Reactors (LWR) under hypothetical accidental conditions such as Loss of Coolant Accident (LOCA), zirconium alloy fuel claddings undergo significant oxidation at high temperatures. To limit this phenomenon, innovative chromium coated nuclear fuel claddings are studied at CEA. However, the integrity under neutron irradiation of such coating for in-service conditions must be preserved..The main objective of this PhD work is to study the behavior under ion irradiation of this new cladding concept. We especially focus on Zr/Cr interface microstructure evolution under irradiation, since the latter controls the adhesion of the coating to the substrate. Ion irradiations were performed to simulate the damage caused by neutrons in LWR. A multi-scale approach is used to characterize the samples before and after irradiation. In particular, Transmission Electron Microscopy (TEM) was used to characterize, at an atomic scale, the microstructure of the Zr/Cr interface. A first type of Zr/Cr interface is observed and shows the presence of nano-metric phases of Zr(Fe,Cr)₂ C14 and ZrCr₂ C15 types. After irradiation, the C14 phase seems to be stabilized over the C15 phase, by segregation of iron at the interface. For a second interface, obtained using different deposition conditions, only C15 phase is observed. Under in-situ TEM irradiation, dissolution of the C15 phase is observed. Whatever the Zr/Cr interface type, preservation of the continuity of crystallographic planes before and after irradiation throughout the interface is demonstrated and thus induces a good adhesion of the coating to the substrate. (author)