Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR

Joyce Conermann, Regis Shogan, Koji Fujimoto, Toshio Yonezawa, Yoichiro Yamaguchi

    Research output: Chapter in Book/Report/Conference proceedingConference contribution

    38 Citations (Scopus)

    Abstract

    Mechanical and corrosion properties were measured on a cold worked, Type 316 stainless steel tube removed from the core of a Pressurized Water Reactor (PWR) after 23 years of service. Neutron exposure levels of the material ranged from near 0 to 65 dpa (∼4.5×1022 n/cm2, E>1 MeV) and irradiation temperatures from 290 to 320°C. As expected, the strength of the material increased and the ductility decreased with irradiation. Slow strain rate and stressed O-ring crack initiation tests in PWR water were used to determine the susceptibility of the material to irradiation assisted stress corrosion cracking (IASCC). The data indicate that IASCC susceptibility may accelerate above 20 dpa and saturate after 60 dpa. At the highest fluence, limited intergranular failure was observed on the fracture surface of a specimen slow strain rate tested in inert gas.

    Original languageEnglish
    Title of host publicationProceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
    Pages277-287
    Number of pages11
    Publication statusPublished - 2005 Dec 1
    Event12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors - Salt Lake City, UT, United States
    Duration: 2005 Aug 142005 Aug 18

    Publication series

    NameProceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors

    Other

    Other12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
    CountryUnited States
    CitySalt Lake City, UT
    Period05/8/1405/8/18

    Keywords

    • IASCC
    • PWR
    • Stainless Steel
    • Swelling

    ASJC Scopus subject areas

    • Engineering(all)

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