A.R. Causey

The effects of intergranular constraints on irradiation growth of Zircaloy-2 at 320 K

Abstract Dimensional changes during neutron irradiation of crystallographically textured Zircaloy-2 has been found to be a complex function of its thermal and mechanical treatments. A model is presented here to investigate the influence of intergranular constraints on the irradiation-induced growth. Good qualitative agreement has been obtained with experimental data for dimensional changes observed in unstressed material, in that, (a) the normal anisotropy of growth, resulting from expansion along 〈 a 〉 axes and contraction along 〈 c 〉 axes can be reversed by prestrain, (b) annealing can introduce intergranular stresses due to the anisotropic thermal expansion, which result in a wider variety of dimensional behaviours than expected from crystallographic texture only, and (c) the intergranular stresses from prestrain are gradually relaxed during irradiation and normal anisotropy eventually returns.

The effect of intergranular stresses on the texture dependence of irradiation growth in zirconium alloys

Abstract Under reactor operating conditions, zirconium alloys have a hexagonal closed-packed crystallographic structure and many of their associated physical properties are anisotropic. Thus the intrinsic deformation processes of plasticity, irradiation creep and irradiation growth result in anisotropic dimensional changes relative to the crystallographic axis in each grain. Since deformation of each grain is constrained by its neighbours, anisotropic deformation generates internal stresses from the intergranular interactions. These stresses cause elastic-plastic and creep deformation whereby each grain conforms to the deformation of its neighbours to maintain the integrity of the sample. The normal anisotropy of growth that results from expansion along the 〈 a 〉 directions and contraction along the 〈 c 〉 directions, cannot be calculated by the ( 1–3f d ) law from simple averaging over the grain orientations, but must take account of the intergranular interactions between grains of differing orientations and the effects of residual stresses. Several models have been developed for the prediction and analysis of irradiation creep and growth in zirconium alloys. In this paper, we review two of these calculations for models based on the self-consistent and upper-bound intergranular constraint theories which illustrate the major effect on intergranular stresses on the texture dependence of irradiation growth.

Anisotropy of in-reactor deformation of Zr–2.5Nb pressure tubes

Theoretical modeling and results from irradiation experiments show that the diametral strain rate of an internally pressurized (closed end) cold-worked Zr-2.5Nb tube increases as the resolved fraction of basal plane normals, f, increases in the radial direction and reduces in the transverse direction. At the same time the elongation rate decreases. The increased diametral strain rate results from an increased diametral creep rate, and a decreased (in magnitude) negative diametral irradiation growth rate. The decreased elongation rate results from a reduction in the axial creep component under the biaxial stress conditions.

In-reactor oxidation of crevices and cracks in cold-worked Zr-2.5 wt% Nb

Abstract Cracks responsible for heavy water leakage from cold-worked Zr-2.5 wt% Nb pressure tubes in units 3 and 4 of Pickering Generating Station were formed by delayed hydrogen cracking. High residual stresses combined with periods when the reactor was cold caused cracks to form the surfaces of which were oxidized during subsequent reactor operation. Analysis of fracture features, oxide thicknesses on crack faces, and the reactor thermal history indicates that the cracks initiated following the first reactor heating cycle. Experiments were done to confirm the time scale deduced from the oxide bands and to explain the presence of thicker oxide in cracks than on adjacent tube surfaces; oxidation in crevices was about six to nine times faster than on free surfaces. The enhancement is attributed to the concentration of LiOH (used to maintain coolant pH) by restricting flow conditions within the crevices.

Measurement of irradiation creep of zirconium alloys using stress relaxation

Abstract Bent-beam stress relaxation tests provide a simple means of assessing the in-reactor creep behaviour of zirconium alloys with different prior thermo-mechanical treatments. By assuming that stress relaxation is equivalent to creep under decreasing stress, the irradiation-induced creep rate can be derived from tests in which small elastically constrained beams are exposed to fast neutron irradiation. In-reactor creep rates for stresses below about 150 MPa have been observed to depend linearly on applied stress, namely ge = Cσφ , hence the stress relaxation behaviour can be described by the relation σ/σ 0 = D exp (−CEφt) where σ/σ 0 is the ratio of the unrelaxed stress (derived from the change in curvature of the unconstrained beam) to the initial stress to bend the beam to the shape of the jig, D describes an initial rapid stress drop, C is a material constant, E is Youngs modulus, φ is fast neutron flux and t is time. We review stress relaxation test results that have been used to characterize the effects of operating parameters (such as temperature, fast neutron flux and fluence) and microstructural properties (such as anisotropy, dislocation structure, grain shape, thermo-mechanical treatments and alloy content) on the in-reactor creep behaviour of zirconium alloys. Correlations between stress relaxation and creep results are also presented with support the use of the relaxation test for gathering in-reactor creep information. Strain recovery in in-reactor and out-reactor annealing experiments suggest that in addition to anelastic strain, a significant part of the in-reactor relaxation strain is due to irradiation damage-related creep.

Thermally induced strain recovery in irradiated and unirradiated zirconium alloy stress-relaxation specimens

Abstract Thermally induced strain recovery has been measured in specimens which were stress relaxed at 570 K either in a reactor or in an autoclave. Strain recovery occurs in both unirradiated and irradiated specimens. Strain recovery in unirradiated specimens is attributed to unpinning of anelastically bowed dislocations. Strain recovery in irradiated specimens occurs in two stages; a rapid stage attributed to unpinning bowed dislocations and a slow stage attributed to annealing of irradiation defects. Assuming that mechanisms proposed for creep are applicable during stress-relaxation the complete recovery of irradiation-induced strains in annealed specimens and partial recovery in specimens with some cold-work agrees with expectations from models based on stress-induced alignment of dislocation loops and irradiation damage. Recovered strains in this experiment were small as were strains measured by others on irradiated specimens with larger deformation strains, thus annealing to reduce strains in reactor structures may not be useful.