Abstract
Studies of the tensile properties of Zircaloy-4 spent fuel cladding and their change with both isothermal and transient annealing have been conducted. The cladding was obtained from spent fuel rods irradiated to a maximum fuel burnup of 30 MWd/kg in the Carolina Power and Light H.B. Robinson power reactor. The yield and ultimate strengths of the as-received material decreased linearly with temperature from room temperature to 42.7/sup 0/C (800/sup 0/F). Uniform elongation was unaffected by temperature over the same range, while total elongation increased sharply between 329 and 371/sup 0/C (625 and 700/sup 0/F). At 482/sup 0/C (900/sup 0/F), properties reflected annealing that occurred during the test. The tensile properties at 371/sup 0/C (700/sup 0/F) were found to be strain rate dependent. The strength properties increased in strain rate, while the total elongation decreased. Uniform elongation exhibited no effect of strain rate. Evidence of dislocation channeling was observed. When the spent fuel cladding was annealed, radiation anneal hardening was noted during early stages in the annealing process. Annealing of irradiation-induced strengthening occurred rapidly at temperatures above 538/sup 0/C (1000/sup 0/F) under isothermal conditions and below 704/sup 0/C (1300/sup 0/F) under transient annealing conditions for heating rates of 28/sup 0/C (50/supmore » 0/F)/s or less. Ductility increases lagged the strength changes during annealing. A ductility minimum, as measured by total elongation, is not reflected in reduction-of-area measurements. The annealing behavior of cold-worked Zircaloy cladding was found to be significantly different from the irradiated material. Annealing was accompanied by a change in the isotropy of deformation as determined from tube wall and diameter measurements. The as-irradiated cladding exhibited essentially isotropic reductions, as opposed to the anisotropic reductions measured for both annealed cladding and unirradiated Zircaloy tubing.« less
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