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.