STP1245

    Corrosion Behavior of Irradiated Zircaloy

    Published: Jan 1994


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    Abstract

    There is ample evidence in the literature of the effects of reactor irradiation on the microstructure and corrosion behavior of zirconium alloys. Specifically, it has been shown that boiling water reactor (BWR) irradiation generally induces nodular corrosion and causes marked changes in precipitate structure and composition. The purpose of this study is to determine the effects of irradiation-induced microstructural changes on post-irradiation corrosion behavior and to gain insight into the operating in-reactor corrosion mechanisms.

    Zircaloy-2 and Zircaloy-4 were irradiated in BWRs at a temperature near 561 K. Neutron fluences were at various values between 2 and 14 × 1025 n/m2 (E > 1 MeV). Post-irradiation corrosion tests were conducted at 589, 673, and 793 K using standard techniques. Transmission electron microscopy (STEM) was conducted on unirradiated, as-irradiated, and corrosion-tested materials.

    Post-irradiation test results include the following: 1. Nodular corrosion as indicated by 793 K steam testing was completely eliminated. Post-irradiation annealing at 923 K/2 h caused a return of nodules to specimen edges. 2. Uniform corrosion as indicated by 589 K water and 673 K steam tests was markedly increased relative to non-irradiated material. (a) The corrosion rate of welded Zircaloy-4 increased rapidly at short times and then achieved a lower steady rate for longer times. (b) The corrosion rate of alpha-annealed Zircaloy-2 and Zircaloy-4 was increased relative to unirradiated material and was linear with time. The corrosion rate was roughly proportional to neutron fluence. Post-irradiation annealing at 923 K/2 h reduced the rate substantially, but not to the unirradiated material value.

    STEM studies of the Zircaloy-4 specimens indicate that the observed corrosion behaviors can be correlated to irradiation-induced increases in the matrix solute concentration. The effects of redistribution and re-precipitation of solute during the corrosion tests are also examined. These results are related to the basic mechanisms of both nodular and uniform corrosion in a BWR.

    Keywords:

    zirconium, nodular corrosion, uniform corrosion, solutes, precipitates, radiation effects, welds, microchemistry, zirconium alloys, nuclear materials, nuclear applications


    Author Information:

    Cheng, B-C
    Principal engineer, senior engineer, and manager, GE Nuclear Energy, Vallecitos Nuclear Center, Pleasanton, CA

    The Electric Power Research Institute, Palo Alto, CA

    Kruger, RM
    Principal engineer, senior engineer, and manager, GE Nuclear Energy, Vallecitos Nuclear Center, Pleasanton, CA

    Adamson, RB
    Principal engineer, senior engineer, and manager, GE Nuclear Energy, Vallecitos Nuclear Center, Pleasanton, CA


    Paper ID: STP15200S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP15200S


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