STP1245: In-Reactor Corrosion Performance of ZIRLO™ and Zircaloy-4

    Sabol, GP
    Consulting engineer, principal engineers, Westinghouse Electric Corporation, Pittsburgh, PA

    Comstock, RJ
    Program manager, Westinghouse Electric Corporation, Science and Technology Center (STC), Pittsburgh, PA

    Weiner, RA
    Consulting engineer, principal engineers, Westinghouse Electric Corporation, Pittsburgh, PA

    Larouere, P
    Staff engineer, Virginia Power, Innsbrook Technical Center, Glen Allen, VA

    Stanutz, RN
    Consulting engineer, principal engineers, Westinghouse Electric Corporation, Pittsburgh, PA

    Pages: 21    Published: Jan 1994


    Abstract

    In-reactor and long-term autoclave corrosion data have been obtained on ZIRLO and three variants of Zircaloy-4: conventional (1.5% tin), low-tin, and beta-treated. In-reactor data from demonstration assemblies irradiated in the Virginia Power Company's North Anna Unit 1 reactor demonstrate the superiority of ZIRLO and, to a lesser extent, low-tin Zircaloy-4 over conventional Zircaloy-4. After two cycles of irradiation to an assembly burnup of 37.8 GWD/MTU, the average axial peak corrosion of ZIRLO was 32% that of conventional Zircaloy-4. Low-tin and beta-treated materials displayed average peak oxides 76% and 150% of that formed on conventional Zircaloy-4, respectively.

    Autoclave corrosion tests of archive tubing have been performed in 633 K water, 672 K, 700 K, and 727 K steam, and in 633 K water containing 70 and 210 ppm lithium as the hydroxide. Correlation of the in-reactor data with the autoclave data indicates that the 633 K pure water test is the best qualitative indicator of in-reactor corrosion performance, and the 672 K steam test the poorest. Differences in in-reactor corrosion between ZIRLO and the Zircaloy-4 materials are consistent with the relative behavior of these materials in lithium hydroxide solutions. The relationships among the in-reactor and autoclave corrosion data, the microstructures, and the processing are discussed.

    In addition to improved corrosion resistance, ZIRLO exhibits improved dimensional stability over Zircaloy-4. The in-reactor creep of ZIRLO is confirmed to be about 80% of that of Zircaloy-4, and irradiation growth is observed to be about 50% that of Zircaloy-4. These data are also presented and discussed.

    Keywords:

    corrosion, zirconium, zirconium alloys, autoclave, water, steam, oxide lithium hydroxide, fuel cladding, processing creep, growth, microstructure, nuclear materials, nuclear applications, radiation effects


    Paper ID: STP15217S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP15217S


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