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    In-Reactor Deformation of Zirconium Alloy Components

    Published: Jan 2010

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    This paper briefly reviews work by the author identifying and describing in-reactor deformation mechanisms of materials and structures used in nuclear reactors, in particular, Zircaloy-2, Zircaloy-4, and Zr-2.5Nb, and the CANDU fuel channel (comprising Zr alloy pressure tubes, calandria tubes, and spacers). The discussion is set in the context of contemporary findings of other workers in the international community The following themes are highlighed: The contributions of creep and growth to deformation; c-component dislocations and the fluence dependence of irradiation growth; anisotrophy of irradiation growth; deformation equations and pressure tube-to-calandria tube contact in CANDU reactors; low temperature flux (damage rate) dependence of deformation rates. The first developments were reported in 1976 at the third conference in this series and there are ongoing developments in all areas. The linear low temperature flux dependence of creep and growth rates is yet to be satisfactorily explained. The original paper was published by ASTM International in the Journal of ASTM International, June 2008.


    irradiation creep, irradiation growth, microstructure, texture, dislocations, anisotropy

    Author Information:

    Holt, R. A.
    Queen's University, Kingston, Ontario

    Committee/Subcommittee: B10.02

    DOI: 10.1520/MNL12125R