Deformation and Fracture Properties of Neutron-Irradiated Recrystallized Zircaloy-2 Cladding under Uniaxial Tension

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.

Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

“…This result is larger than the 570-MPa yield strength reported for the as-received material in the annealed condition and is greater than most values reported in the literature for unirradiated Zircaloy-4. [5][6][7][8][15][16][17] However, it is worth noting that in a study using burst tests, Yagnik et al 18 reported hoop strength similar to those measured here for room temperature tests of unirradiated Zircaloy-4. A more detailed analysis of cladding deformation will be addressed in future efforts.…”

“…The angled fracture surface is consistent with previously reported results for room temperature testing of irradiated Zircaloy. 6 The high-magnification image (upper left of Fig. 46) reveals a dimpled fracture surface typical of ductile fracture.…”

“…46) reveals a dimpled fracture surface typical of ductile fracture. 6,20 Figures 49 and 50 display series of SEM micrographs showing two fracture surfaces from a Zircaloy-4 specimen irradiated to 50-GWd/MT fuel burnup. Unlike the fracture from the 21-GWd/MT burnup specimen shown in Fig.…”

“…[5][6][7] Higgy and Hamad 5 reported results for tensile testing of irradiated Zircaloy-2 and Zircaloy-4. Specimens were irradiated at low temperature (<100°C) up to a fluence of 1.43 × 10 20 neutrons/cm 2 (>1 MeV) and at elevated temperature (320-360°C) up to 1.53 × 10 21 neutrons/cm 2 (>1 MeV).…”

“…Yasuda et al 6 irradiated Zircaloy-2 at ~290°C and performed tensile testing at varying strain rates and temperatures. Test temperatures from 25°C to 400°C and strain rates from 0.05 to 5% per minute were applied.…”