High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods under Dry Storage Conditions

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Abstract
Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571°C in limited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn’s analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding, which decreases the internal rod pressure. The cladding creep also contributes to radial cracks, through the external oxide and internal fuel-cladding chemical interaction layers, which propagated into and arrested in an oxygen stabilized alpha-Zircaloy layer. There were no signs of either additional cladding hydriding, stress corrosion cracking, or fuel pellet degradation. If irradiation hardening does not reduce the stress rupture properties of Zircaloy, a conservative maximum storage temperature of 400°C based on a stress-rupture mechanism is recommended to ensure a 1000-yr cladding lifetime.