STP633: Rupture Characteristics of Zircaloy-4 Cladding with Internal and External Simulation of Reactor Heating

    Fiveland, WA
    Senior research engineers, Babcock & Wilcox Company, Alliance Research Center, Ohio

    Barber, AR
    Senior research engineers, Babcock & Wilcox Company, Alliance Research Center, Ohio

    Lowe, AL
    Principal engineer, Babcock & Wilcox Company Nuclear Power Generation Division, Lynchburg, Va.

    Pages: 14    Published: Jan 1977


    Abstract

    Rupture characteristics of internally heated Zircaloy-4 nuclear fuel cladding were investigated to develop rupture data that would provide a basis to bench-mark mathematical modeling of clad rupture. Single- and five-tube configurations of pressurized water reactor cladding were ruptured at constant pressures and heating rates with simulated reactor boundary conditions. The results showed that the α + β material property transformation region had a strong effect on the material ductility, and the clad rupture temperature varied linearly with hoop stress in the α + β region. Axial distribution of multitube ruptures was not coplanar. The rupture positions on all specimens were distributed normally about the point of maximum temperature. This paper emphasizes test techniques as well as the rupture characteristics of Zircaloy-4 clad.

    Keywords:

    zirconium, zirconium alloys, nuclear fuel claddings, rupture characteristics


    Paper ID: STP35563S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35563S


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