STP683: Estimates of Time-Dependent Fatigue Behavior of Type 316 Stainless Steel Subject to Irradiation Damage in Fast Breeder and Fusion Power Reactor Systems

    Brinkman, CR
    Group leader and research staff members, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Liu, KC
    Group leader and research staff members, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Grossbeck, ML
    Group leader and research staff members, Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Pages: 21    Published: Jan 1979


    Abstract

    Cyclic lives obtained from strain-controlled fatigue tests at 593°C of specimens irradiated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63 × 1026 neutrons (n)/m2 (E > 0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20 percent cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradiations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadings ranging from 2 to 5 MW/m2 were used. Results, although conjectural because of the many assumptions, tended to show that 20 percent cold-worked Type 316 stainless steel could be used as a first-wall material meeting a 7.5 to 8.5 MW-year/m2 lifetime goal provided the neutron wall loading does not exceed more than about 2 MW/m2. These results were obtained for an air environment, and it is expected that the actual vacuum environment will extend lifetime beyond 10 MW-year/m2.

    Keywords:

    irradiation, stainless steel, elevated temperature, fatigue, creep-fatigue interaction, ductility, fusion, fission, reactor


    Paper ID: STP38184S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP38184S


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