STP1450: Preliminary Results of the United States Nuclear Regulatory Commission's Pressurized Thermal Shock Rule Reevaluation Project

    Dickson, TL
    Oak Ridge National Laboratory, Oak Ridge, TN

    Williams, PT
    Oak Ridge National Laboratory, Oak Ridge, TN

    Bass, BR
    Oak Ridge National Laboratory, Oak Ridge, TN

    Kirk, MT
    Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission, Washington, DC

    Pages: 16    Published: Jan 2004


    Abstract

    The current federal regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early to mid-1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of an improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license-extension considerations. Based on the above, the United States Nuclear Regulatory Commission (USNRC) initiated in 1999 a program to re-evaluate the current PTS regulations within the framework established by modem probabilistic risk assessment techniques.

    As part of the USNRC PTS project, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, human reliability analysis, materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. The experts have been from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions.

    Recently, the FAVOR code was applied to a domestic commercial pressurized water reactor to evaluate the adequacy of the current regulations and to determine if a technical basis can be established to support a change in the current regulations. This paper gives an overview of the improved computational methodology and presents some results of the preliminary analyses.

    Keywords:

    fracture mechanics, pressurized thermal shock, probabilistic, Monte Carlo


    Paper ID: STP11286S

    Committee/Subcommittee: E08.08

    DOI: 10.1520/STP11286S


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