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.
Nuclear plant reactor pressure vessel heat-up and cool-down pressure temperature (P-T) limit curves are determined using ASME Section XI, Appendix G. ASME has adopted into ASME Section XI Appendix G the allowable use of the ASTM E1921 master curve fracture toughness based reference temperature (T0) to index the KIc curve. ASME Section XI Code Case N-830 allows the use of the KJc 95% lower bound master curve indexed using T0 directly. In ASME Section XI Appendix G, the equation RTT0 = T0 + 19.4°C, as an alternate RTNDT, shifts the KIc curve to approximate the KJc 95% lower bound master curve, however, the KIc exponential curve parameters are different. Thus, this paper evaluates the impact on plant heatup and cooldown pressure-temperature limit curves between the two ASME approved methods for typical pressurized water reactors (PWR). Different degrees of embrittlement are assessed to determine differences in the two approaches on reactor pressure vessel (RPV) beltline operating curves. Furthermore, in the proposed Revision 19 of Regulatory Guide 1.147, the US NRC has included a condition on the use of Code Case N-830 that prohibits the use of the current KIc equation in ASME Section XI Appendix G when these values are above the KJc lower bound 95% curve at temperatures below T0 − 64°C. This paper briefly discusses this NRC condition on the P-T limit curves.
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.
A recent increase in operating experience (OE) related to pipe cracking in non-isolable auxiliary piping systems has been realized in the Pressurized Water Reactor (PWR) nuclear power industry. The majority of PWR auxiliary piping systems are comprised of welded stainless steel pipe and piping components. The susceptible piping systems are Class 1 pressure boundary and typically non-isolable from the primary loop. Since they are non-isolable, when a pipe crack or crack indication is identified, an emergent flaw evaluation and/or repair is required. Typically, the evaluations begin with an ASME Section XI IWB-3640 flaw evaluation to determine acceptability of the as-found flaw at the time of shutdown. Subsequent flaw evaluations are performed to demonstrate the possibility of continued operation of the piping component by leaving the flaw as-is without repair. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth evaluations, and the detection capabilities of the non-destructive examination technique. If evaluation of the as-found indication does not produce acceptable results, then a repair/replacement activity per ASME Section XI is considered. Possible repair scenarios include replacement of the piping section or component, or structural weld overlay. The results of the flaw evaluations or repairs must ensure that the auxiliary piping system will continue to operate safely. This paper will discuss the recent experiences of two different piping systems (boron injection tank line and drain line) that experienced cracking, the potential causes for the cracking in the absence of evidence, and the ASME Code flaw evaluations and/or repairs performed to support continued operation of the plant.
Since the implementation of pressure-temperature (P-T) limit curves in the 1960s for light water reactors, the P-T limit curves have been based on the limiting locations in the reactor coolant system, which are typically the irradiated reactor pressure vessel (RPV) region adjacent to the core (beltline) and the closure head flange. Recently, it has been questioned as to whether the reactor vessel inlet or outlet nozzle corners could be more limiting due to the stress concentration at these locations. The discussion presented in this paper provides technical justification that the RPV nozzle corner P-T limit curves are bounded by the traditional P-T limit curves for the pressurized water reactors (PWRs). The current approach in evaluating the Pressurized Water Reactor Inlet and Outlet nozzle corner regions with respect to plant heatup and cooldown Pressure Temperature Limit Curves contains a number of conservatisms. These conservatisms include postulation of a large 1/4T flaw at the nozzle corner region, use of RTNDT (reference nil-ductility temperature), and fracture toughness prediction based on plane strain fracture toughness. The paper herein discusses several factors that can be considered to improve the pressure temperature limit curves for nozzle corners and increase the operating window for nuclear power plant operations. Prior to the 2013 edition, the ASME Section XI Appendix G did not require the use of a 1/4T flaw for the nozzle corners; furthermore, a smaller postulated flaw size is permissible. Based on inspection capability and experience, a smaller flaw size can easily be justified. The use of a smaller flaw size reduces the stress intensity factors and allows for the benefit of being able to take advantage of increased material toughness due to the loss of constraint at the nozzle corner geometry. The analysis herein considers the calculation of stress intensity factors for small postulated nozzle corner flaws based on a 3D finite element analysis for Westinghouse PWR inlet and outlet nozzle corner regions. The Finite Element Analysis (FEA) stress intensity factors along the crack front are used in the determination of allowable pressures for the cooldown transient Pressure-Temperature limit curves.
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