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Zircaloy cladding embrittlement was evaluated from data derived from the postirradiation examination of 47 light-water reactor type fuel rods irradiated under postulated accident conditions. The previously irradiated and unirradiated rods were tested in 15 separate in-pile experiments in the Power Burst Facility reactor at the Idaho National Engineering Laboratory as part of the Thermal Fuels Behavior Program. Film boiling conditions ranged in duration from a few seconds to about 15 min, with cladding peak temperatures ranging from 1315 to ≥ 2100 K in the most severely oxidized Zircaloy. Quenching from film boiling temperatures through the β → α′ phase transformation was at rates of about 50 to 100 K/s, depending on test conditions.
The Zircaloy-4 cladding showed rapid oxygenation of the β-phase Zircaloy and hydriding in excess of the pretest characterization. Cladding embrittlement in the intact rods was consistent with previously established oxygen embrittlement criteria. Fuel rods operated with a cladding defect were embrittled to a greater extent than intact rods oxidized under similar conditions. A modified background prior β-phase and precipitation of rim-α was typically observed in the failure region of breached rods. The fracture resistance of breached rods appears to be related to enhanced hydrogen absorption and its influence on both the β-phase transformation and the α′-morphology during rapid (∼ 100 K/s) quenching.
nuclear industry, zirconium, zirconium alloys, Zircaloy-4, embrittlement, hydrogen, oxidation, nuclear fuel rods, irradiation
Scientist, LWR Safety Research Section, LWR Fuel Research Division, EG&G Idaho, Inc., Idaho Falls, Idaho