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    Corrosion of Zircaloy in the Presence of LiOH

    Published: 01 January 1991

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    The corrosion of Zircaloy-4 is accelerated in water or steam above about 350°C when LiOH is present in sufficient concentrations. A short-term test (16 h) to predict inreactor corrosion behavior has been developed based on this accelerated corrosion. A group of claddings with a range of in-reactor corrosion behavior has been tested in LiOH solutions from 310 to 415°C with Li concentrations up to 375 ppm. Up to 100 times the rate seen in pure water was obtained in supercritical water with high LiOH levels. The best correlation to the in-reactor behavior was obtained for a Li content in the range of 75 to 190 ppm in the temperature range of 390 to 410°C.

    The corrosion rate for Zircaloy-4 in steam containing LiOH has been observed to be higher than in water with LiOH for tests below the critical temperature. For tests above the critical temperature, the corrosion rate was greater for samples which were above the initial water line in the capsule. These samples would have been exposed to LiOH-bearing steam during the initial heat up of the samples. Therefore sufficient Li is present in the steam phase to cause accelerated corrosion.

    The accelerated test with LiOH correlated well with 415°C/3 day autoclave test results for Zircaloy-4 but not for zirconium-based alloys containing niobium. The 415°C/3 day test has been found to correlate well to in-reactor corrosion performance for Zircaloy-4. Based on the assumption that this test is also effective for evaluating the corrosion resistance of other zirconium-based alloys, the accelerated LiOH test appears to be effective only within the Zircaloy-4 composition range. Therefore the LiOH accelerated test should not be used for evaluating new alloy compositions.


    uniform corrosion, LiOH, supercritical water, Zircaloy, cladding, strip, inreactor correlation

    Author Information:

    Perkins, RA
    SIEMENS Nuclear Power Corporation, Richland, WA

    Busch, RA
    SIEMENS Nuclear Power Corporation, Richland, WA

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP25529S