No abstract

No abstract

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian Deuterium Uranium Pressurized Heavy-Water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm2 hydrogen per year). The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes.

Zr-2.5Nb alloy pressure tubes for CANDU® reactors are nominally extruded at 815°C, cold-worked about 27%, and stress-relieved at 400°C for 24 h. The resulting structure consists of elongated α-Zr grains interspersed with a network of thin β-Zr filaments. Corrosion tests on unirradiated and preirradiated material have investigated the effects of microstructure and microchemistry on corrosion and hydrogen ingress. In two-phase (α-Zr+β-Zr) structures, the corrosion and hydrogen pickup increases with increasing volume fraction of β-Zr. Corrosion is highest for single β-phase material although hydrogen pickup reverts to a minimum value. Tests on alloys with low Nb concentration show that the optimum corrosion resistance occurs at a Nb content of about 0.1 wt% Nb. Thermal aging the metastable two-phase structure reduces corrosion and is consistent with a lower β-phase volume fraction and a lower concentration of Nb in the α-phase. Cold working the as-extruded two-phase structure up to about 80%, prior to stress relieving, reduces the out-reactor corrosion by about a factor of two. However, in-reactor, the benefits of cold work are negligible since there is a suppression of corrosion in irradiated Zr-2.5Nb that dominates all other effects. Irradiation results in an increase in dislocation density due to dislocation loop formation and also enhances the progression to an equilibrium α-phase composition manifested by the appearance of Nb-rich precipitates. Both of these effects of irradiation on microstructure are associated with improved corrosion properties based on tests of materials with controlled microstructures and microchemistry. Any thermally induced decomposition of the α-phase, resulting from the stress-relief heat-treatment, is slowed or even reversed by irradiation, depending on flux and temperature, and is therefore unlikely to have a significant effect on corrosion of irradiated materials. One of the most important factors leading to improved corrosion properties in Zr-2.5Nb pressure tubing seems to be the precipitation of β-Nb particles and the concomitant reduction of Nb in the matrix of the α-Zr grains during irradiation. Apart from any direct effects of cold-working or dislocation loop formation, it is likely that increased dislocation densities will also enhance Nb precipitation.

X-ray diffraction (XRD) and analytical electron microscopy (AEM) have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in PHWR reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and, in addition, improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For PWR and BWR reactors, the onset of “breakaway” growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure.