The behavior of Zircaloy-clad fuel rods for light water reactors under off-normal operating conditions is being studied in the Power Burst Facility. Fuel rods from three irradiation experiments conducted under power-cooling-mismatch conditions have been examined destructively in remote handling facilities. The postirradiation examinations of these pressurized water reactor-type rods have included visual and photographic examination, physical dimensioning, and metallography.

High-temperature oxidation and mechanical strength under a loss-of-coolant accident (LOCA) were investigated using irradiated and unirradiated fuel cladding. Cladding from fuel rods irradiated up to 49 GWD/MTU in a Japanese commercial PWR and unirradiated cladding, most of which was preoxidized and precharged with hydrogen to simulate high burnup fuel, were subjected to the tests. High-temperature oxidation tests showed that the oxidation weight gain for irradiated cladding was equal to, or slightly lower than, that for unirradiated material. Preoxidized cladding showed less oxidation weight gain, and no effect of hydrogen absorption on oxidation behavior was observed. The mechanical tests (uniaxial strength and ductility) after a thermal sequence simulating a LOCA showed comparable behavior with that of unirradiated cladding, due to the recovery of the irradiated microstructure. These test results suggest: (1) there is no adverse effect of irradiation on the high-temperature oxidation behavior; and (2) radiation damage in cladding is eliminated during a LOCA condition.