No abstract

Transient-heating tube-burst tests were conducted on a 50.8-cm length of Zircaloy cladding obtained from spent-fuel rods irradiated in the H. B. Robinson power reactor to a peak burn-up of 30 MWD/kg. Internal electrical resistance heaters were used to achieve nominal heating rates of 28°C/s for the tests. The tests were conducted in steam and the independent experimental variable was the initial level of helium pressurization. Tube burst pressures varied from 0.965 to 12.514 MPa. Corresponding burst stresses and temperatures varied from 7.5 MPa and 1135°C to 99.1 MPa and 734°C, respectively. Failure strains ranged up to 46 percent when burst occurred at upper (α + β) and β-phase temperatures and up to 30 percent for failures occurring at lower temperatures. No significant difference in burst behavior between unirradiated and spent nuclear fuel Zircaloy cladding was observed experimentally. No influence of axial restraint on burst strains was noted.

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No abstract

The objective of this program was to provide a mechanical-property data base that could be used to predict the performance of Zircaloy-4 clad fuel rods under various power reactor conditions. The data were developed for off-normal and transient reactor conditions where the source of cladding loads included thermal stresses and internal gas pressure. Irradiated fuel-rod cladding was obtained from the Westinghouse H. B. Robinson Reactor and the Babcock and Wilcox Oconee I Reactor. The irradiated fuel rods were first characterized using visual examination, gamma scanning, spiral profilometry, and eddy current tests to provide data for assessing the general condition of the as-received fuel rods and for determining the suitability of the cladding for testing. A group of specimens was transient annealed at heating rates of 0.56, 5.6, 14, and 28°C/s to maximum temperatures ranging from 482 to 816°C. After annealing, the specimens were tested at 371°C and at a strain rate of 0.004/min. The burst properties for both the Oconee I and the H. B. Robinson fuel-rod cladding exhibited similar trends after transient anneals. Large decreases in burst stresses and increases in burst strains occurred for specimens transient annealed above 600°C for both Oconee I and H. B. Robinson cladding. The strain recovery tended to lag the stress decreases for transient annealed burst specimens for both reactor materials as the maximum annealing temperature was increased.