Hydride reorientation and its impact on mechanical properties of high burn-up and unirradiated cold-worked stress-relieved Zircaloy-4 and ZirloTM fuel cladding

Precipitation of radial hydrides in zirconium-based cladding concomitant with the cooling of spent nuclear fuel can potentially compromise cladding integrity during its subsequent handling and transportation. Complementary experimental programs on unirradiated and irradiated cladding were performed to assess the conditions leading to the precipitation of radial hydrides and their detrimental effects on mechanical properties. This sequel paper to a previous publication [1] presents results for cold-worked, stress-relieved Zircaloy-4 and ZirloTM cladding used in pressurized water reactors. Cold-worked, stress-relieved Zircaloy-4 and ZirloTM behave very similarly, although irradiated ZirloTM was found to be slightly more prone to hydride reorientation (HRO) than irradiated Zircaloy-4. No distinct irradiation-induced effect on the HRO critical stress was detected. Depending on the initial high temperature that varied from 350 to 450 °C, HRO may cause a ductile-to-brittle transition when the cladding hoop stress, created by internal pressurization of the specimen, ranges from 70 MPa to 120 MPa. Cracking of the oxide layer caused local preferential radial hydride precipitation. The impact of the decreasing temperature and internal pressure during cool-down was also investigated: it has been found that samples undergoing decreasing temperature and pressure cycles are less prone to hydride reorientation.