STP1366: Void-Induced Swelling and Embrittlement in the Russian EI-847 Stainless Steel at PWR-Relevant End-of-Life Conditions

    Garner, FA
    Senior Staff Scientist, Pacific Northwest National Laboratory, Richland, WA

    Porollo, SI
    Head of Hot Laboratory, Lead Scientist, Head of Nuclear Materials Division, Leading Scientist and Senior Scientist, Institute of Physics and Power Engineering, Obninsk,

    Vorobjev, AN
    Head of Hot Laboratory, Lead Scientist, Head of Nuclear Materials Division, Leading Scientist and Senior Scientist, Institute of Physics and Power Engineering, Obninsk,

    Konobeev, YV
    Head of Hot Laboratory, Lead Scientist, Head of Nuclear Materials Division, Leading Scientist and Senior Scientist, Institute of Physics and Power Engineering, Obninsk,

    Dvoriashin, AM
    Head of Hot Laboratory, Lead Scientist, Head of Nuclear Materials Division, Leading Scientist and Senior Scientist, Institute of Physics and Power Engineering, Obninsk,

    Krigan, VM
    Head of Hot Laboratory, Lead Scientist, Head of Nuclear Materials Division, Leading Scientist and Senior Scientist, Institute of Physics and Power Engineering, Obninsk,

    Budylkin, NI
    Leading Scientist, Senior Scientist, Bochvar's Institute of Nonorganic Materials, Moscow,

    Mironova, EG
    Leading Scientist, Senior Scientist, Bochvar's Institute of Nonorganic Materials, Moscow,

    Pages: 10    Published: Jan 2000


    Abstract

    Void swelling has been proposed to eventually reach significant levels in some austenitic internals at PWR end-of-life conditions. A prevailing perception based on previously published data from fast reactors is that swelling will never reach significant levels at temperatures in the 330–372°C range, even at the 80–90 dpa level anticipated in PWR baffle-former components after 40 years of operation. Fast reactor data are presented in this paper that show that this perception is incorrect, at least for a niobium-stabilized 16Cr-15Ni steel used in Russian reactors for applications where 316 would be used in Western reactors. As a result of the large swelling levels reached in some cases, the steel becomes very brittle, as expected from fast reactor experience.

    Keywords:

    stainless steels, void swelling, embrittlement, PWRs, neutron irradiation


    Paper ID: STP12438S

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP12438S


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