Microstructure of zirconium fuel claddings: TEM and EBSD studies of as-received and ne-utron-irradiated materials

Abstract

Zirconium fuel claddings act as a first barrier against release of fission products during nuclear power plant operation and interim storage of the spent fuel. During the reactor operation, cladding tubes are exposed to different stress level at elevated temperatures and neutron irradiation in corrosive environment. It causes a material degradation by corrosion, cladding embrittlement by hydrides and radiation-induced damage or radi-ation growth and creep of the fuel rods. The irradiation damage effects mainly contribute to the loss of material ductility. In our study, microstructure of as-received (non-irradiated) Zr-alloys used in LWR (Zr1Nb, Zr-1Nb-1.2Sn-0.1Fe, Zr-1.5Sn-0.2Fe-0.1Cr) were examined by electron microscopy methods. Transmission electron microscope (TEM) was used to describe the microstructure of claddings used in different reactor conditions and identify the radiation-induced damage, which is presented on Zr1Nb irradiated to one standard campaign in the VVER-1000 active zone. Following Electron Backscatter Diffraction (EBSD) method on transparent foils complements the TEM results in larger area, i. e. by grain size and orientation or analysis of local misorienta-tion after irradiation. Radiation-induced damage was observed in Zr1Nb metallic matrix as type disloca-tion loops, presence of radiation-induced precipitates or partial amorphization of the secondary phase particles. EBSD method showed no changes in crystallographic orientation, but a local increase of dislocation density can be affected by neutron irradiation.