As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.
One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis. The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.
Nuclear Regulatory Commission (NRC) guidance for Subsequent License Renewal (SLR) applicants is provided in the NUREG-2191. Specifically, Section X.M1 of the NUREG-2191 includes guidance for aging management programs (AMP) which has two aspects related to fatigue analyses, one that verifies the continued acceptability of existing analyses through cycle counting and the other that provides periodically updated evaluations of the fatigue analyses to demonstrate that they continue to meet the appropriate limits. In addition, the NUREG-2191 provides requirements to account for reactor water environment by deriving cumulative fatigue usage (CUF) including environmental effects (CUFen) in component fatigue analyses for a set of sample critical components for the plant outlined in NUREG/CR-6260. Furthermore each applicant is required to determine plant-specific component locations in the reactor coolant pressure boundary that may be more limiting than those considered in NUREG/CR–6260. This paper presents a methodology to identify plant-specific primary equipment component locations that are potentially more limiting than the locations identified in NUREG/CR-6260, through a detailed review and ranking of the analyses of record (AORs) for a plant. The ranking approach implemented in this methodology assesses the level of technical rigor required to derive the CUF values and provides a means to appropriately compare CUF values within the same transient section. This paper also illustrates how this methodology was applied for a recent license renewal application.
As part of a fatigue management program for subsequent license renewal, a flaw tolerance evaluation based on ASME Code, Section XI, Appendix L may be performed. The current ASME Code, Section XI, Appendix L flaw tolerance methodology requires determination of the flaw aspect ratio for initial flaw size calculation. The flaw aspect ratios listed in ASME Section XI, Appendix L, Table L-3210-2, for austenitic piping for example, are listed as a function of the membrane-to-gradient cyclic stress ratio. The Code does not explicitly describe how to determine the ratio, especially when utilizing complex finite element analyses (FEA), involving different loading conditions (i.e. thermal transients, piping loads, pressure, etc.). The intent of the paper is to describe the methods being employed to determine the membrane-to-gradient cyclic stress ratios, and the corresponding flaw aspect ratios (a/l) listed in Table L-3210-2, when using finite element analysis methodology. Included will be a sample Appendix L evaluation, using finite element analysis of a pressurized water reactor (PWR) pressurizer surge line, including crack growth calculations for circumferential flaws in stainless steel piping. Based on this example, it has been demonstrated that, unless correctly separated, the membrane-to-gradient cyclic stress ratios can result in extremely long initial flaw lengths, and correspondingly short crack growth durations.
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