Hydraulic Burst Tests at Elevated Temperatures on Zircaloy Cladding from Fuel Rods Irradiated in the Winfrith SGHWR.

摘要

Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to < = 15.3 MWd/kgU, equivalent to an integrated fast neutron flux < = 3.4 x 10 exp 25 n/m exp 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallized conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallized condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallized condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallized Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (Atomindex citation 12:585891)