The in-reactor corrosion of Zircaloy-2 and Zr-2.5Nb pressure tube materials is being studied by exposing small specimens in water-cooled loops in the NRU research reactor. Experiments are designed to investigate the effects of fast-neutron irradiation, water chemistry, temperature, and pre-oxidation on the corrosion behaviour. Preliminary results show that inreactor corrosion rates of both alloys increase with increasing concentrations of dissolved oxygen in the water. In the absence of irradiation, only Zr-2.5Nb is sensitive to its presence. Excursions to oxidizing (⟩300 (μg/kg oxygen) water chemistry in nominally low oxygen tests are seen to induce high oxidation rates for both Zircaloy-2 and Zr-2.5Nb during irradiation. The consequences of oxygen excursions are less pronounced on material prefilmed in 673 K steam. The oxides formed in different chemistry regimes are characterized using scanning electron microscopy and capacitance techniques.
Over the temperature range 528 to 568 K, the temperature dependence for corrosion of pickled specimens during irradiation is low with activation energies being marginally higher for Zr-2.5Nb than for Zircaloy-2. In these tests it is seen that fast-neutron irradiation increases corrosion of Zircaloy-2, whereas for Zr-2.5Nb a decrease in corrosion is frequently observed during irradiation.