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 radiationinduced damage or radiation 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 misorientation after irradiation. Radiationinduced damage was observed in Zr1Nb metallic matrix as <a> type dislocation 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.

Effect of heat-treatment on material properties of L-DED printed austenistic alloy 08CH18N10T for nuclear reactor applications

“…Metallographical analysis of the microstructure was performed using a light optical microscopy (LOM) and scanning electron microscopy (SEM). For further analysis of grain structures, electron back-scatter diffraction mapping (EBSD) was also used [9]. Metallography was performed on heat treated and non-heat treated L-DED fabricated 08CH18N10T to observe differences in microstructure.…”

ATF Claddings after High-Temperature Steam Oxidation: WDS and Nanoindentation Studies to Characterize in-depth Material Changes

“…Due to a good balance between resistance to corrosion, neutron transparency and mechanical properties, zirconium-based alloys have been used since the 1960s for nuclear fuel claddings in Light Water Reactors (LWRs) [1]. Preserving the good mechanical properties and integrity of the fuel cladding is crucial not only for the reactor operation, but also for the following wet and dry storage and transportation of the spent fuel rods [2]. In addition, this role becomes more important during loss of coolant (LOCA) and severe accidents [3].…”