Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    Volume 4, Issue 7 (July 2007)

    ISSN: 1546-962X

    CODEN: JAIOAD

    Published Online: 9 August 2007

    Page Count: 24


    Wang, J. A.
    Oak Ridge National Laboratory, Oak Ridge,

    Rao, N. S. V.
    Oak Ridge National Laboratory, Oak Ridge,

    Konduri, S.
    Oak Ridge National Laboratory, Oak Ridge,

    (Received 5 June 2007; accepted 17 July 2007)

    Abstract

    A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5 % and 52 % in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.


    Paper ID: JAI100681

    DOI: 10.1520/JAI100681

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    Title Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels
    Symposium 23rd Symposium on Effects of Radiation on Materials, 2006-06-15
    Committee E10