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    Multiaxial In-Reactor Stress-Rupture Strength of Stainless Steels and a Nickel Alloy

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    In the framework of the cladding program for fast breeders we have tested the behavior of tubes under constant pressure during irradiation. The test facility and the type of specimens are described briefly.

    The maximum available helium pressure was 500 atm and the test temperature between 600 and 700 C. Tests were run with the stainless steels, 16Cr-13Ni, 20Cr-25Ni, and Incoloy 800, and a nickel alloy, Inconel 625. The irradiation was done in the BR2 reactor at Mol, Belgium, with a fast neutron flux of 2 to 4 × 1014 n/cm2 s and a maximum time-to-rupture of 3000 h. Identical tests were done without irradiation, and the results were as follows: 16Cr-13Ni—The decrease of the stress-rupture strength under irradiation was in the range of 40 to 50 percent and the absolute tangential strain between 0.4 and 1.0 percent. Inconel 625—The comparison of different types of unirradiated specimens showed the influence of the specimen dimensions and test conditions. The stress-rupture strength of irradiated tubes was decreased by 30 to 40 percent; the corresponding tangential strain was smaller than 1 percent. Incoloy 800—The outstanding result of this alloy was the relatively high tangential strain under irradiation (3.0 to 4.5 percent). The reduction of stress-rupture strength by irradiation was of the same order as for the stainless steel. 20Cr-25Ni—The reduction of strength upon irradiation was in the range of 30 percent, and the tangential strain was smaller than 1 percent. The rupture time at a load of 9.5 kg/mm 2 at 600 C was only 2500 h.

    The metallographic tests of all the irradiated specimens have shown in each case small intergranular cracks and no burst. The results are discussed and conclusions are reached concerning the optimal fuel-element design.


    stress rupture strength, stainless steels, nickel alloys, irradiation, tangential rupture strain, ductility, intergranular cracks, secondary creep stage, creep-strength, high-temperature embrittlement

    Author Information:

    Laue, H. J.
    Dr.-Ing., Institut für Angewandte Reaktorphysik, Kernforschungszentrum Karlsruhe,

    Böhm, H.
    Professor, Dr.-Ing., Institut für Material- und Festkorperförschung, Kernforschungszentrum Karlsruhe,

    Hauck, H.
    Dr.-Ing., Institut für Material- und Festkörperforschung, Kernforschungszentrum Karlsruhe,

    Committee/Subcommittee: E10.02

    DOI: 10.1520/STP41859S