STP754

    Burst Criterion of Zircaloy Fuel Claddings in a Loss-of-Coolant Accident

    Published: Jan 1982


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    Abstract

    A burst criterion model has been developed to calculate the burst data of Zircaloy claddings of a pressurized water reactor in a loss-of-coolant accident. It is assumed by this model that the time of burst is reached when the local stress equals the limiting burst stress. The burst stress is assumed to depend on the temperature and oxygen concentration of the Zircaloy.

    With the histories of temperature and pressure known, the strain, stress, and oxygen content can be calculated by integration of the creep equation and correlation of the oxidation kinetics. Once the time of burst has been calculated, all burst data, that is, strain, stress, temperature, pressure, and oxygen content, are determined.

    To verify the burst criterion model, single-rod transient burst tests in steam were performed using fuel rod simulators with indirect electric heating. The test parameters covered the following ranges: internal overpressure—10 to 140 bar, heating rate—1 to 30 K/s.

    The burst temperatures and burst strains predicted by the burst criterion model are in good agreement with the test data.

    Keywords:

    zirconium alloys, nuclear industry, nuclear fuel cladding, plastic deformation, oxidation, pressurized water reactors, loss-of-coolant accident, burst criterion model, burst tests, burst temperature, burst stress, burst strain


    Author Information:

    Erbacher, FJ
    Kernforschungszentrum Karlsruhe, Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit, Karisruhe,

    Neitzel, HJ
    Kernforschungszentrum Karlsruhe, Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit, Karisruhe,

    Rosinger, H
    Atomic Energy of Canada Limited, Whiteshell Nuclear Research Establishment, Pinawa, Manitoba

    Schmidt, H
    Kernforschungszentrum Karlsruhe, Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit, Karisruhe,

    Wiehr, K
    Kernforschungszentrum Karlsruhe, Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit, Karisruhe,


    Paper ID: STP37058S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP37058S


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