No abstract

Sufficient evaluation of the changes in mechanical properties, such as elastic, plastic, and failure properties, due to neutron irradiation in service is required to precisely predict fuel performance. This paper presents the results of the uniaxial tensile tests performed for recrystallized (850 K, 2.5 h) Zircaloy-2 claddings irradiated in commercial BWRs to fluences of 5 × 1023 to 4 × 1025 n/m2 (E > 1 MeV). The material constants of irradiated Zircaloy-2 were obtained precisely, using a high temperature elongation detector in a hot-cell and computer analyses of digital stress-strain data. The tensile tests were carried out at 298 to 673 K at strain rates of 0.05 to 5%/min, using tubular specimens cut to 135 mm lengths from the claddings. From these tests, Young's modulus, strain hardening exponent, strain rate sensitivity, and fracture behavior were obtained and evaluated. Young's modulus of specimens irradiated to 1024 n/m2 is nearly equal to that of unirradiated ones, but irradiation to 1025 n/m2 shows values 7% higher than for unirradiated ones. The irradiation-induced increments in yield stress are rapid below 1 × 1024 n/m2, then slow down above this fluence becoming proportional to (φt)0.1 for fluences of 1 × 1024 to 4 × 1025 n/m2, where φt is fluence. The strain hardening exponent before irradiation is 0.15 and a constant which is independent of strain levels, but after irradiation it decreases depending on strain increase. The strain rate sensitivity at the 0.2% plastic strain decreases from 0.037 to 0.021 by irradiation. There are two types of fracture morphologies after irradiation: necking and spiral types. In the latter, a shear band propagates along only one direction. Both are accompanied by localized deformation bands caused by dislocation channelling. The spiral type is predominant at about 573 K, in high irradiation and at high strain rate. The fracture surfaces of the unirradiated and irradiated materials are all ductile, while the number of small dimples, whose nuclei appear to be irradiation defects, increases with fluence. The ratio of the number for unirradiated, irradiation to 1 × 1024 n/m2, and irradiation to 1 × 1025 n/m2 is 1:2.7:2.9.

A series of measurements was performed to examine the effects of texture, hydrogen concentration, thermal cycling and morphology of hydrides on stress reorientation of hydrides using unirradiated recrystallized Zircaloy-2 sheet materials. It was found that the threshold stress for reorientation of hydrides is approximately 80 MPa, and the threshold stress is not affected by hydrogen concentration, thermal cycling or hydride morphology before the hydride reorientation test. Over the threshold stress, texture, thermal cycling and morphology of hydrides affect the stress reorientation. The modified Ells' equation was proposed for expressing the effect of the texture of Zircaloy-2.

Knoop hardness was measured in order to study plastic anisotropy of irradiated fuel cladding tubes. Results are summarized as follows. The lowest Knoop hardness was obtained when the indenter was so oriented that (1010} (1210) prism slip played the dominant role in indentation formation. The dependence of Knoop hardness on neutron fluence implied that the damage caused by neutron irradiation affected the prism slip of a-zirconium more strongly. The considerable amount of plastic anisotropy observed in the unirradiated tubes showed a tendency to decrease with increasing neutron fluence.

Initial plastic deformation behavior of zirconium alloy fuel cladding was described quantitatively by the deformation system of single crystal of a-zirconium, and a model was proposed to simulate the yield behavior of polycrystalline material. Based on the model, effects of crystallographic texture and stress state on the plastic deformation of the cladding were evaluated. Conclusions obtained from this investigation are :( 1) The proposed model shows good agreement with the von Mises' yield criteria for a material with isotropic properties. (2) Plastic anisotropy of the cladding decreases when neutron irradiation affects prism slip more strongly than the other deformation systems.(3) Dominant deformation systems for axial tension or internal pressurization of the cladding are predicted to be prism slip or tensile twin, respectively, when the stress state of the cladding reaches the yield condition.