STP1423

    The Influence of In-Situ Clad Straining on the Corrosion of Zircaloy in a PWR Water Environment

    Published: Jan 2002


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    Abstract

    Zircaloy fuel element cladding changes dimensions during service in a pressurized water reactor (PWR) as a result of stress-free irradiation-induced growth, creep-driven by fuel pellet expansion and hydriding. The application of a tensile load in a high-temperature autoclave environments has previously been reported to increase the corrosion rate of Zircaloy, and heat treatments (beta quenching) that reduce the irradiation-induced stress-free growth of Zircaloy have previously been reported to reduce Zircaloy corrosion in-reactor. However, the effect of in-situ straining on Zircaloy corrosion in a PWR environment has not been systematically studied and reported in the literature. This paper presents experimental results regarding the effect of in-situ straining on Zircaloy corrosion in a PWR environment, both in-reactor and in an autoclave. In-situ electrochemical data and post-test metallographic data are presented. In-situ straining is seen to increase the corrosion rate of the Zircaloy, likely by breaking the passivating layer that is forming on the surface. However, the effect is a function of the applied strain rate.

    Keywords:

    Zircaloy, in-situ clad straining, corrosion


    Author Information:

    Kammenzind, BF
    Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA

    Eklund, KL
    Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA

    Bajaj, R
    Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA


    Paper ID: STP11405S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP11405S


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