You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.


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

    Published: 0

      Format Pages Price  
    PDF (1.3M) 38 $25   ADD TO CART
    Complete Source PDF (18M) 843 $121   ADD TO CART


    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

    Author Information:

    Erbacher, FJ
    Institut für Reaktorbauelemente, Projekt Nukleare Sicherheit,

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

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP28138S