High temperature post-irradiation performance of spent pressurized-water-reactor fuel rods under dry-storage conditions

Post-irradiation studies on failure mechanisms of well characterized PWR rods were conducted for up to a year at 482, 510 and 571/sup 0/C in unlimited 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 pressures. The cladding creep also contributes to radial cracks, through the external oxide and internal FCCI layers, which propagated into and arrested in an oxygen stabilized ..cap alpha..-Zircaloy layer. There were no signs of either additional cladding hydriding, stress-corrosion cracking or fuel pellet degradation. Using the Larson-Miller formulization, a conservative maximum storage temperature of 400/sup 0/C is recommended to ensure a 1000-year cladding lifetime. This accounts for crack propagation and assumes annealing of the irradiation-hardened cladding

Spent fuel resistance to internally produced cladding degradation

These tests were conducted over a narrow temperature range considerably above anticipated disposal conditions and utilized only one set of rods from a single reactor. The following conclusions are made: The measured cladding strain was sufficiently large so that failure mechanism verification by inducing breaches in unmodified rods heated to elevated temperatures for short periods of time does not appear to be practical based on a stress rupture mechanism. At the elevated test temperatures, though, Blackburn's formulization based on stress rupture gives very conservative estimates of breach times. In addition to the high cladding strain, the fuel exhibited no additional gas release or axial fission product migration at 482/sup 0/C. The nondestructive examination gave no additional indication of internal deterioration of the fuel rod

Oxidation of spent fuel in air at 175 degree to 195 degree C

Oxidation tests in dry air were conducted on four LWR spent fuels at 175{degrees} and 195{degrees}C to determine the effect of the fuel characteristics on the oxidation state likely to exist at the time leaching occurs in a potential repository. Weight changes were measured and samples were examined by XRD, ceramography, TEM, and TGA. Despite local variations in the grain boundary susceptibility to oxidation, all four fuels progressed toward an apparent endpoint at an oxygen-to-metal (O/M) ratio of 2.4. The sole oxidation product was U{sub 4}O{sub 9+x,} a cubic phase structurally related to UO{sub 2} but with a slightly smaller lattice constant. The growth of the U{sub 4}O{sub 9+x} from the grain boundaries into the UO{sub 2} grains followed parabolic kinetics and had an activation energy of 26.6 kcal/mol. Based on the kinetics, the time required at 95{degrees}C to completely oxidize LWR spent fuel to U{sub 4}O{sub 9+x} would be at least 2000 yr. The next oxidation product to form after the U{sub 4}O{sub 9+x} phase may be U{sub 3}O{sub 8,} but no U{sub 3}O{sub 8} or other dilatational oxidation product has been detected in these accelerated tests conducted up to 25,000 h