STP633

    Corrosion Monitoring of Steam Generating Heavy Water Reactor Pressure Tubes

    Published: Jan 1977


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    Abstract

    The corrosion of zirconium alloys under boiling water reactor (BWR) conditions is enhanced by irradiation; therefore, it was necessary to verify the expected behavior in steam generating heavy water reactor (SGHWR) of pressure tabes manufactured by different routes. Specimens of each material and others of interest were exposed to the coolant inside perforated fuel cans replacing fuel pins in otherwise standard SGHWR fuel clusters. After irradiation, specimens were decrudded chemically and weight gains were measured. Hydrogen pick-up was found by hot vacuum degassing. Specimens were also examined metallographically.

    Weight gains from irradiations up to 531 effective full-power days (EFPD) generally show a reasonably linear correlation with the fast fluence to which they have been exposed. There is a variation in corrosion rate between Zircaloy-2 materials of different histories; a quench treatment at the billet stage seems to be beneficial. Metallographic examination of specimens and fuel cladding showed the presence of nodular corrosion after exposure ia reactor in oxygenated water. The extent of nodular formation varied with material and generally followed the weight gain patterns. The paper includes a discussion on mechanistic aspects of nodular corrosion.

    Hydrogen up-take by the specimens appears to be largely independent of corrosion weight gain making percentage pick-up figures of dubious value. The amounts of hydrogen taken-up, however, are small, representing less than 5 percent for many materials. Extrapolation of results to the expected full reactor life indicates that neither loss of section by corrosion nor hydrogen pick-up by the pressure tube should be life limiting.

    Keywords:

    zirconium, heavy water reactors, zirconium alloys, pressure vessels, corrosion, nodules


    Author Information:

    Sheppard, MF
    Section leader and head, United Kingdom Atomic Energy Authority, Risley Engineering and Materials Laboratory, Risley, Warrington,

    Tyzack, C
    Section leader and head, United Kingdom Atomic Energy Authority, Risley Engineering and Materials Laboratory, Risley, Warrington,


    Paper ID: STP35575S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35575S


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