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Localized oxidation of zirconium alloys in high temperature and pressure oxidizing environments of nuclear reactors



This paper presents the results of parametric studies on the existing test procedures to assess nodular corrosion susceptibility of zirconium (Zr) base alloys. The parameters included the level of dissolved oxygen (DO) in the steam, exposure time, and type of exposure. The alloys studied were Zircaloy-2, Zircaloy-4, Zr–2.5Nb, and Zr–1Nb. A two-step test procedure involving prefilming at 410 °C followed by nodule growth at 510 °C in deaerated steam (using demineralized water) was found to be the most representative for Zircaloys and none of the test conditions produced nodules on Zr–Nb alloys. Presence of high dissolved oxygen was found to suppress nodule formation in Zircaloys. A detailed investigation on the morphology of individual nodules and nodule cross section is presented. Mechanism of nodular corrosion, in particular and localized oxidation of Zr alloys at 400 °C, in general, in oxidizing environments of nuclear reactors has been discussed in the light of the present results of nodular corrosion. Long term oxidation of Zircaloy-2 and Zircaloy-4 under low and high DO conditions are reported and compared with our earlier results on the effect of dissolved oxygen on the long term oxidation and hydrogen pick-up behavior of Zr–Nb alloys.