STP1023: A Systematic Survey of the Factors Affecting Zircaloy Nodular Corrosion

    Ogata, K
    Nippon Nuclear Fuel Development Co., Ltd., Ibaraki,

    Yoshitsugu, M
    Professor Emeritus, University of Tokyo, Tokyo,

    Okubo, T
    Academic vice-president and professor, Sophia University, Tokyo,

    Aoki, T
    Nuclear Power Engineering Test Center, Tokyo,

    Hattori, T
    The Tokyo Electric Power Co., Inc., Tokyo,

    Fujibayashi, T
    Isogo Engineering Center, Toshiba Corp., Yokohama,,

    Inagaki, M
    Hitachi Research Laboratory, Hitachi, Ltd., Hitachi,

    Murota, K
    Japan Nuclear Fuel Co., Ltd., Yokosuka,

    Kodama, T
    Sumitomo Metal Industries, Ltd., Tokyo,

    Abe, K
    Materials Research Laboratories, Kobe Steel, Ltd., Kobe,

    Pages: 24    Published: Jan 1989


    Abstract

    Out-of-reactor corrosion tests in 10.3-MPa steam at temperatures between 490 and 530°C have been carried out on Zircaloy-2 tube specimens of various fabrication histories including archive specimens of the same tube lots as were used in the commercial boiling water reactor (BWR) to get data for comparison between in- and out-of-reactor nodular corrosion performances. Test conditions, such as temperature and duration, as well as material parameters, such as chemical composition, texture, and precipitates, were examined to clarify the relation to the nodular corrosion susceptibility.

    The number density and size of nodules increased when tested at higher temperatures. Extending the test duration increased the nodule size, but not the number density. The archive tube lots showed different degrees of nodular corrosion in the out-of-reactor test and their relative rankings seemed to correlate with the visual ratings of in-reactor corrosion. A higher number density of precipitates tended to reduce the nodular corrosion susceptibility. It should be noted, however, that some lots constituted an exception to this, and the cause was sought in the tube manufacturing processes.

    Keywords:

    zircaloy, steam tests, in-reactor nodular corrosion, boiling water reactor, fabrication histories


    Paper ID: STP18871S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP18871S


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