STP939: Zircaloy Fuel Cladding Behavior in a Loss-of-Coolant Accident: A Review

    Erbacher, FJ
    Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit,

    Leistikow, S
    Institut für Material- und Festkörperforschung, Projekt Nukleare Sicherheit,

    Pages: 38    Published: Jan 1987


    This paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for the emergency core cooling (ECC) in a loss-of-coolant accident (LOCA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation, and coolability of deformed rod bundles.

    The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS). Reference is made to test data obtained in other countries.

    The paper concludes that the major mechanisms and consequences of oxidation, deformation, and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC criteria are conservative and that the coolability of a LWR and the public safety in a LOCA can be maintained.


    zirconium alloys, pressurized water reactors, loss-of-coolant accident, oxidation, embrittlement, plastic deformation, cooling

    Paper ID: STP28138S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP28138S

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