Effect of the accelerated irradiation and hydrogen/helium gas on IASCC characteristics for highly irradiated austenitic stainless steels

K. Fujimoto, Toshio Yonezawa, E. Wachi, Y. Yamaguchi, M. Nakano, R. P. Shogan, J. P. Massoud, T. R. Mager

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

    33 Citations (Scopus)

    Abstract

    In order to clarify irradiation assisted stress corrosion cracking (IASCC) characteristics of highly irradiated austenitic stainless steels over dose of 40dpa, SCC susceptibility in simulated PWR primary water, mechanical/ metallurgical properties, and hydrogen/helium gas concentrations have been investigated for the austenitic stainless steels irradiated in actual PWRs or in fast breeder reactor (FBR). The IASCC susceptibility of the stainless steels irradiated in a FBR was extremely lower than that of the stainless steels irradiated in a PWR, according to the slow strain rate tensile (SSRT) tests in PWR environment. Mechanical properties and radiation induced segregation (RIS) of the stainless steels irradiated in a FBR revealed the same tendency as those of the stainless steels irradiated in a PWR. But hydrogen/helium gas concentrations in the stainless steels irradiated in a FBR was extremely smaller than that in the stainless steels irradiated in a PWR. Also, cavity formation was observed not only in the grain, but also near the grain boundary in the highly irradiated stainless steel in a PWR. Therefore, it is suggested that the hydrogen/helium gas plays an important role in IASCC of PWR core internals, in addition to conventionally proposed RIS and radiation hardening. Furthermore, it is considered that the IASCC characteristics at the high irradiation range are different from those at the low irradiation range.

    Original languageEnglish
    Title of host publicationProceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
    Pages299-310
    Number of pages12
    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

    • Baffle Former Bolt
    • Fast Breeder Reactor
    • Helium
    • Hydrogen
    • Intergranular Stress Corrosion Cracking
    • Irradiation Assisted Stress Corrosion Cracking
    • Pressurized Water Reactor
    • Radiation Induced Segregation

    ASJC Scopus subject areas

    • Engineering(all)

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