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    STP1572

    Comparison of Charpy V-Notch-Energy-Based Embrittlement Trend Curves Developed in the United States

    Published: 2014


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    Abstract

    Charpy V-notch energy (CVE) versus temperature data obtained from the U.S. light water nuclear reactor fleet are used to track irradiation damage to reactor pressure vessel (RPV) materials. This data is typically fit using a hyperbolic tangent (tanh) function to determine the temperature at which the mean CVE is equal to 30 ft-lb (41 J); the information used in tracking irradiation embrittlement. EricksonKirk [EricksonKirk, M. T., “Technical Basis for Revision of Regulatory Guide 1.99: NRC Guidance on Methods to Estimate the Effects of Radiation Embrittlement on the Charpy V-Notch Impact Toughness of Reactor Vessel Materials,” ADAMS ML081120289, U.S. NRC Draft Report, U.S. Nuclear Regulatory Commission, Washington, D.C., 2007] and others recently developed an alternative strategy for fitting the CVE versus temperature data in which they identified a single CVE versus temperature relationship that appears to well represent the behavior of a very wide variety of ferritic steels for temperatures at and below the fracture mode transition temperature. This Charpy “master curve” equation was developed in a manner consistent with that used to define a fracture toughness transition reference temperature as described in ASTM E1921 [1] with specimens exhibiting significant ductility excluded. The newly defined TCVE was used to develop an embrittlement trend curve (FIT 6) to predict irradiation damage in RPV steels and was demonstrated to predict that data reasonably well. Predictions of embrittlement behavior made using this embrittlement trend curve (ETC) were recently compared to predictions made using other trend curves developed using the U.S. light water reactor fleet [“Radiation Embrittlement of Reactor Vessel Materials,” Regulatory Guide 1.99, Revision 2, U.S. Nuclear Regulatory Commission, Washington, D.C., 1988; ASTM E900-02: Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, Annual Book of ASTM Standards, ASTM International, West Conshohocken, PA, 2007; and Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” NRC Regulation 10CFR50.61a, U.S. Nuclear Regulatory Commission, Washington, D.C., 2010]. Comparisons were made using the entire data set as well as on subsets of the data (BWR versus PWR data, specific product forms, high copper versus low copper, etc.). These comparisons demonstrate that although Fit 6, 10CFR50.61a, and ASTM E900 all predict the data fairly well, some curves are better at predicting various aspects of the data than others. The strengths and weaknesses of each curve are discussed in detail.


    Author Information:

    Erickson, Marjorie
    Phoenix Engineering Associates, Inc., Unity, NH


    Committee/Subcommittee: E10.02

    DOI: 10.1520/STP157220130124