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    Investigation of Materials from a Decommissioned Reactor Pressure Vessel—A Contribution to the Understanding of Irradiation Embrittlement

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    For the first time, material samples from the reactor pressure vessel of a decommissioned nuclear power plant (Gundremmingen unit A, Federal Republic of Germany, FRG) were trepanned and investigated in the FRG within an international cooperative program to establish material toughness data in comparison with those obtained from the surveillance program and the initial material state. The material under consideration is base material from several vessel shells, circumferential weld and heat affected zone. The available data on this material include acceptance data, irradiated state data, and data obtained from the thermally aged only condition. The fluence of the vessel wall was in the range of 3–4 · 1018 cm−2. The material response to neutron radiation is compared with the Code trend curves.

    For the material of one vessel shell, severe degradation (an upper-shelf drop down to 55 J) was found in the transverse direction, which is not in accordance with the existing trend curves. This fact was not recognized from the surveillance program, since, according to the practice used in the past, specimens of this type were not included. Annealing of those specimens did not provide the expected recovery. The behavior of a second vessel shell and the weld material could conservatively be predicted by the Code. From the available data it must be concluded that the high degradation of the one vessel shell is typical for only this specific material, and cannot be attributed to the in-service parameters of the reactor. Irradiation experiments of archive material in test reactors, however, did not indicate this severe susceptibility to neutron radiation.


    irradiation embrittlement, ferritic material, surveillance program, pressure vessel, postirradiation heat treatment, Charpy impact testing, directionality of forgings

    Author Information:

    Kussmaul, Karl
    Prof. Dr.-Ing, Universität Stuttgart,

    Föhl, Jürgen
    Dr.-Ing, Universität Stuttgart,

    Weissenberg, Thomas
    Dipl.-Ing, Universität Stuttgart,

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP49445S