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Closed-end burst tests were conducted on sections of fuel cladding cut from five Pickering-type fuel bundles after exposures of 2.5 to 4.8 × 1024 n/m2 (>1 MeV) at 300 C (572 F). Five batches or heat treatments of Zircaloy-2 and -4 were tested at 22 and 300 C (72 and 572 F). Two of the bundles contained an unfueled element in the center ring. The burst test results obtained from fueled and unfueled sheaths are compared with the same cladding batches irradiated to similar fluences without fuel in a fast neutron facility at 125 to 250 C (257 and 482 F) and to unirradiated controls.
The fueled cladding specimens exhibited higher strength and lower total circumferential elongation than unfueled specimens irradiated in the fast neutron facility. Although the fueled cladding was strained as much as 1.2 percent by thermal expansion of the UO2, the uniform elongation in subsequent burst tests was often higher than that obtained on bursting tubing irradiated in the fast neutron facility. Most of these differences were attributed to the different irradiation temperatures that produced a variation in the size distribution of radiation damage and perhaps a variation in the stability of the defect loops under strain.
While there was no visual or metallographic evidence of interaction between the fuel and the cladding, there was a slight difference in properties between the fueled cladding and unfueled sheaths contained in the same fuel bundles. The slightly lower strength and total circumferential elongation of the fueled cladding were taken as evidence of some mechanical interaction, probably the 0.1 to 1.2 percent plastic strain imposed on the sheath by the thermal expansion of the UO2. There were indications of a texture effect in irradiation hardening with the greater values occurring in instances where deformation took place almost entirely by slip.
Hydrogen concentrations in the range 170 to 719 ppm drastically reduced the 22 C (72 F) strength and ductility, but the 300 C (572 F) properties were significantly reduced only in cladding with a low level of cold work.
neutron irradiation, fast neutrons, radiation effects, nuclear fuel cladding, tubing, zirconium alloys, Zircaloys, uranium dioxide, mechanical properties, burst test, ductility, elongation, tensile strength, yield strength, fractures (materials), metallography, twinning, cold working, heat treatment, electron microscopy, texture, zirconium hydrides, strains, slip, lattice defects, dislocations
Group leader, Atomic Energy of Canada Ltd., Chalk River, Ontario