STP458: Corrosion and Failure Characteristics of Zirconium Alloys in High-Pressure Steam in the Temperature Range 400 to 500 C

    Johnson, A. B.
    Senior Research Scientist, Chemistry Research Department, Pacific Northwest Laboratory, Battelle Memorial Institute, Richland, Wash.

    Pages: 15    Published: Jan 1969


    A nuclear fuel-failure phenomenon prompted a study of the corrosion properties of the Zircaloy-2 fuel cladding. The principal studies were conducted at 400, 450, 475, and 500 C in steam at 1700 psig. Materials tested were: five lots of Zircaloy-2; one lot of Ozhennite (Zr-0.2Sn-0.1Fe-0.1Ni-0.1Nb); and one lot of crystal bar zirconium. As-fabricated (rolled or extruded) Zircaloy-2 underwent a rapid approach to failure in tests at 475 and 500 C. The attack involved nucleation of nodular oxide which spread across the surface, finally resulting in an accelerated, relatively uniform attack. Ozhennite was resistant to the attack. Heat treatments (α and β anneals) delayed the catastrophic attack on Zircaloy-2 but not on crystal bar zirconium.


    zirconium alloys, catastrophic corrosion, high-temperature corrosion, fuel cladding

    Paper ID: STP43833S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP43833S

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