Microstructural Changes of Fuel Claddings for Pressurized Water Reactors Fuel after Irradiation

摘要

Microstructural changes of spent fuel claddings discharged from pressurized water reactors were investigated by a radiation-shielded micro-X-ray diffraction system (n-XRD). A zircaloy-4 cladding of UO2 fuel and a zirlo cladding of burnable poison fuel (5.98 wt % gadolinia-urania) with 41 GWd/tU bum-up and a zircaloy-4 cladding of 58 GWd/tU UO2 fuel were chosen as sample specimens to investigate the effects of burn-up on the irradiation behaviour of cladding. At 41 GWd/tU bum-up, the chemical interaction between a fuel pellet and cladding was not observed for both samples. The formation of ZrO2 and ZrH2 phases was observed in the outer surface of the zircaloy-4 cladding, while those phases were not observed for the zirlo cladding. It can be concluded that zirlo is more stable than zircaloy-4 at this bum-up. At 58 GWd/tU bum-up, the formation of a zirconium oxide (ZrOx) layer was observed at the inner surface of the zircaloy-4 cladding, which indicates that the cladding interacts with the fuel. On the outer surface of zircaloy-4 cladding, ZrO2 and ZrH2 phases were also observed.