The terminal solid solubility of hydrogen in zirconium alloys

experimental program has been carried our to measure the terminal solid solubility (TSS) of hydrogen in Zircaloy. Unirradiated electron beam (EB) welded Zircaloy-2, EB-welded Zircaloy-4, Zircaloy-2 pressure tube, cold-worked P-quenched Zircaloy-2 forging, oxygen-strengthened tungsten inert gas (TIG) welded Zircaloy-4, and irradiated welded Zircaloy-2, welded Zircaloy-4, and Zircaloy-2 pressure tube materials were tested. A differential scanning calorimetry technique was used to measure the dissolution and precipitation temperatures for a range of hydrogen concentrations. The TSS behavior of unirradiated EB welded Zircaloy 2 and Zircaloy-4, Zircaloy-2 pressure tube, and P-quenched Zircaloy-2 forging materials was very similar, indicating little influence of chemical composition or microstructure on TSS. There was a marked hysteresis between the dissolution (TSSd) and precipitation (TSSp) temperatures, and best fit equations are provided for the two curves. Oxygen-strengthened TIG welded Zircaloy-4 exhibited markedly different solubility behavior to the other unirradiated materials. The oxygen addition increased the hydrogen solubility. Two explanations have been postulated for the effect, either an increase in the matrix strength or hydrogen trapping at the solute atoms. However, increasing the matrix strength in Zircaloy-2 by cold working did not increase hydrogen solubility. Zircaloy-2 and Zircaloy-4 materials, irradiated to fluences in the range 5.5 X 10(20) to 1.0 x 10(22) n/cm(2) (E > 1 MeV) at temperatures in the range of 250 to 300 degrees C, have higher dissolution and precipitation solubilities compared to those measured on unirradiated material. When the irradiation damage was annealed out at 500 degrees C, the TSS temperatures tended to be restored to the unirradiated values. It has been hypothesized that the effect of irradiation is to trap hydrogen at the irradiation damage sites, although a second explanation involving the increase in matrix strength due to irradiation has also been considered. For unirradiated Zircaloy, thermal history was found to affect TSSp but not TSSd. Increasing the peak temperature or the hold period at peak temperature reduced the TSSp temperature. This effect is considered due to a reduction of the "memory effect" in which preferential sites for hydride precipitation are removed by the annealing. Heating and cooling rates over the range 0.5 to 10 degrees C/min had little effect on the measured dissolution and precipitation temperatures.