STP1046V2: Experimental Assessments of Gundremmingen RPV Archival Material for Fluence Rate Effects Studies

    Hawthorne, J. Russell
    Senior metallurgist, Materials Engineering Associates, Inc., Lanham, MD

    Hiser, Allen L.
    Engineer, Materials Engineering Associates, Inc., Lanham, MD

    Pages: 25    Published: Jan 1990


    Abstract

    The 250-MW, boiling-water Gundremmingen reactor KRB-A in the Federal Republic of Germany (FRG) has been decommissioned. A joint USA/FRG/UK study, conceived by the U.S. Nuclear Regulatory Commission (NRC), is underway to evaluate material removed from the vessel for a critical assessment of power reactor versus test reactor environment effects. The vessel operated at ∼288°C; the inner-wall fluence estimate at decommissioning was about 1×1019n/cm2, E>1MeV.

    This report describes test reactor irradiation assessments of a forging segment believed to be archive material from the KRB-A vessel fabrication. Charpy-V (Cv), compact tension (CT), and tension test specimens were evaluated in five as-irradiated conditions and two postirradiation annealed conditions. With 288°C irradiations, the elevation in 100 MPa · m1/2 temperature was found to match the elevation in 41-J temperature within 12°C. The latter elevation was predicted well by NRC's Regulatory Guide 1.99.

    The L-C orientation data for the archive material versus the vessel trepans suggest a fluence-rate effect. However, the C-L orientation data do not. A test orientation dependence of radiation embrittlement sensitivity, described by the trepan material, but not the archive material, is responsible and is anomalous.

    Keywords:

    Gundremmingen KRB-A reactor, pressure vessels, radiation embrittlement, notch ductility, fracture toughness, neutron dosimetry, 20NiMoCr26 steel, A336 steel, test reactor, Charpy-V test, compact tension test, postirradiation annealing, J-R curve, fluence rate effects


    Paper ID: STP49444S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP49444S


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