STP633: Embrittlement of Zircaloy-4 by Liquid Cesium at 300°C

    Syrett, BC
    Senior metallurgist, senior scientist, and metallurgy program manager, Stanford Research Institute, Menlo Park, Calif

    Cubicciotti, D
    Senior metallurgist, senior scientist, and metallurgy program manager, Stanford Research Institute, Menlo Park, Calif

    Jones, RL
    Senior metallurgist, senior scientist, and metallurgy program manager, Stanford Research Institute, Menlo Park, Calif

    Pages: 14    Published: Jan 1977


    Abstract

    Previous attempts to induce embrittlement of zirconium alloys by cesium at light water power reactor operating temperatures (300 to 350°C) have been unsuccessful, although embrittlement has been observed in the 30 to 225°C temperature range. Thus, although cesium is an abundant fission product, it has been discounted previously as a possible cause of failures of Zircaloy-clad fuel rods. In the present study, reactor-grade Zircaloy-4 tube specimens suffered embrittlement in molten cesium at 300°C as well as at 40°C. However, it appeared that embrittlement occurred only if the cesium had a low oxygen content and was contaminated with a trace of iron.

    Keywords:

    zirconium, zirconium alloys, Zircaloy R, liquid metals, cesium, embrittlement, stress corrosion, failure


    Paper ID: STP35576S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35576S


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