Mark Kirk

DEVELOPMENT OF THE ALTERNATE PRESSURIZED THERMAL SHOCK RULE (10 CFR 50.61a) IN THE UNITED STATES

In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRCs efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.

Review of Proposed Methodology for Risk-Informed Relaxation to ASME Section XI: Appendix G

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.Copyright

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

The current federal regulations to ensure 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 that were 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 the 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, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (F racture A nalysis of V essels: O ak R idge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.Copyright

Mechanistic Insights Into Risk-Informed Revision of ASME Section XI–Appendix G

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity, throughout their operating life, when subjected to planned normal reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are generally considered to be conservative and some plants are finding it operationally difficult to heat-up and cool-down within the accepted limits. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to increase operational flexibility while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) are reviewing the industry proposed risk-informed methodology. Previous results of this review, have been reported at PVP, and a NRC report summarizing all results is currently in preparation. The objective of this paper is to discuss and illustrate mechanistic insights into trends shown previously associated with normal cool-down.Copyright

Potential roles for the Master Curve in regulatory application

The Master Curve, as introduced by Wallin and co-workers in 1984, has evolved into a mature technology for characterizing the fracture toughness transition of ferritic steels. Considerable empirical evidence provides testament to the robustness of the Master Curve procedure. However, in 1997, the Nuclear Regulatory Commission (NRC) staff detailed several technical issues requiring resolution prior to staff acceptance of applications of Master Curve technology to the fracture integrity assessment of nuclear reactor pressure vessels (RPVs) [1]. Current and recently completed research programs sponsored by both the NRC and Electric Power Research Institute (EPRI) focus on closure of these issues. This paper reviews the issues detailed in 1997, comments on their continued relevance in light of recent research results, and details areas where either additional research or a change or research focus is warranted.

Effect of Cladding Residual Stress Modeling Technique on Shallow Flaw Stress Intensity Factor in a Reactor Pressure Vessel

Prior probabilistic fracture mechanics (PFM) analysis of reactor pressure vessels (RPVs) subjected to normal cool-down transients has shown that shallow, internal surface-breaking flaws dominate the RPV failure probability. This outcome is caused by the additional crack driving force generated near the clad interface due to the mismatch in coefficient of thermal expansion (CTE) between the cladding and base material, which elevates the thermally induced stresses. The CTE contribution decreases rapidly away from the cladding, making this effect negligible for deeper flaws. The probabilistic fracture mechanics code FAVOR (Fracture Analysis of Vessels, Oak Ridge) uses a stress-free temperature model to account for residual stresses in the RPV wall due to the cladding application process. This paper uses finite element analysis to compare the stresses and stress intensity factor during a cool-down transient for two cases: (1) the existing stress-free temperature model adopted for use in FAVOR, and (2) directly applied RPV residual stresses obtained from empirical measurements made at room temperature. It was found that for a linear elastic fracture mechanics analysis, the application of measured room temperature stresses resulted in a 10% decrease in the peak stress intensity factor during a cool-down transient as compared to the stress-free temperature model.Copyright

Derivation of the New Pressurized Thermal Shock Screening Criteria

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.Copyright

Models for Use in Probabilistic Pipe Rupture Assessment Code Development

US Nuclear Regulatory Commission (USNRC) Standard Review Plan (SRP) 3.6.3, describes the current methodology for leak-before-break (LBB) piping safety assessment. Specifically, it describes a deterministic assessment procedure that can be used to demonstrate compliance with the 10CFR50 Appendix-A, General Design Criterion 4 (GDC-4) requirement that the primary system pressure piping exhibit an extremely low probability of rupture. However, SRP 3.6.3 does not permit assessment of piping systems with active degradation mechanisms, even though it is known that primary water stress corrosion cracking (PWSCC) has occurred in systems that have been granted LBB exemptions to remove pipe-whip restraints. To address this need, a program is being conducted with the long-term goal of developing a probabilistic assessment tool that can be used to directly demonstrate compliance with 10CFR50 Appendix–A, GDC-4, a tool that would account for the effects of both active degradation mechanisms and the mitigation activities that are being undertaken to address this degradation. This program has been termed “xLPR” as its goal is to demonstrate an eX tremely L ow P robability of R upture in pressure boundary piping systems. This methodology augments current LBB assessment models (leak rate and crack stability models) through the addition of best estimate models describing the initiation and propagation of flaws due to the various degradation mechanisms (fatigue, PWSCC, intergranular stress corrosion cracking (IGSCC), etc.), inspection models, and mechanical and chemical mitigation/remediation models that describe changes in stress state, pipe material and environment caused by mitigation/remediation efforts. Models currently used in LBB assessment will be updated, or replaced, with best estimate, probabilistic models, including those for leak rate and crack stability assessment. All models should account for the full distribution of input variables (where known) in order to account for both epistemic and aleatory uncertainties in as detailed a manner as feasible. This paper summarizes the structure and current activities of the Modeling Task Group within the framework of the overall xLPR Project, and the methodology used to select and develop the models for the xLPR Pilot Study. Preliminary information on the various models chosen for the Pilot Study, and how they are linked within the structure of the overall xLPR probabilistic code, is also provided.Copyright

Assessment of the Temperature Dependence of Ferritic Steel Fracture Toughness on or Near the Lower Shelf

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case N-830. This revision incorporates a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all the fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J-R) from the lower shelf through the upper shelf. A paper presented at the 2014 ASME Pressure Vessel and Piping Conference described most of these models; a companion paper at this conference describes the J-R model. This paper also supports the WGFE effort by performing an assessment of the appropriateness of Wallin’s Master Curve model to represent toughness data on the lower shelf, and by comparing the Master Curve with the current Code KIc curve on the lower shelf.The work presented in this paper supports the following conclusions:1. The Master Curve provides a reasonable representation of cleavage fracture toughness (KJc) data at lower shelf temperatures. A statistical evaluation of a large database demonstrates that the Master Curve works well to temperatures approximately 140 °C below To or, equivalently, approximately 160 °C below RTTo.2. The percentile of cleavage fracture toughness data falling below a KIc curve indexed to RTTo varies considerably with temperature. At lower shelf temperatures as much as half of the data lie below the KIc curve, while at temperatures close to RTTo this percentage falls to approximately ≈ 1.5%. The current guidance of Nonmandatory Appendix A to Section XI to use structural factors of √10 or √2 is one means of addressing this inconsistency.3. The inconsistent degree to which the KIc curve, with or without structural factors, bounds fracture toughness data cannot be fixed within the current Code framework for two reasons: the KIc curve does not reflect the actual temperature dependence shown by the fracture toughness of ferritic RPV steels, and the ratio of a mean or median toughness curve to a fixed percentile bound is not a constant value. It is for these reasons that in the next revision of Code Case N-830 the ASME WGFE is moving away from use of the KIc curve coupled with structural factors and, instead, is adopting models of fracture toughness that represent both the temperature trends and the scatter in the data with high accuracy.Copyright

Options for Defining the Upper Shelf Transition Temperature (Tc) for Ferritic Pressure Vessel Steels

This paper evaluates current guidance concerning conditions under which the analyst is advised to transition from a linear-elastic fracture mechanics (LEFM) based analysis to an elastic-plastic fracture mechanics (EPFM) based analysis of pressure vessel steels. Current guidance concerning the upper-temperature (T>c) for LEFM-based analysis can be found in ASME Section XI Code Case N-749. Also, while not explicitly stated, an upper-limit on the KIc value that may be used in LEFM-based evaluations is sometimes taken to be 220 MPa√m (a value herein referred to as KLIM). Evaluations of Tc and KLIM were performed using a recently compiled collection of toughness models that are being considered for incorporation into a revision to ASME Section XI Code Case N-830; those models provide a complete definition of all toughness metrics needed to characterize ferritic steel behavior from lower shelf to upper shelf. Based on these evaluations, new definitions of Tc and KLIM are proposed that are fully consistent with the proposed revisions to Code Case N-830 and, thereby, with the underlying fracture toughness data. Formulas that quantify the following values over the ranges of RTTo and RTNDT characteristic of ferritic RPV steels are proposed:• For Tc, two values, Tc(LOWER) and Tc(UPPER), are defined that bound the temperature range over which the fracture behavior of ferritic RPV steels transitions from brittle to ductile. Below Tc(LOWER), LEFM analysis is acceptable while above Tc(UPPER) EPFM analysis is recommended. Between Tc(LOWER) and Tc(UPPER), the analyst is encouraged to consider EPFM analysis because within this temperature range the competition of the fracture mode combined with the details of a particular analysis suggest that the decision concerning the type of analysis is best made on a case-by-case basis.• For KLIM, two values, KLIM(LOWER) and KLIM(UPPER), are defined that bound the range of applied-K over which ductile tearing will begin to occur. At applied-K values below KLIM(LOWER), ductile tearing is highly unlikely, so the use of the KIc curve is appropriate. At applied-K values above KLIM(UPPER), considerable ductile tearing is expected, so the use of the KIc curve is not appropriate. At applied-K values in between KLIM(LOWER) and KLIM(UPPER), some ductile tearing can be expected, so it is recommended to give consideration to the possible effects of ductile tearing as they may impact the situation being analyzed.These definitions of Tc and KLIM better communicate important information concerning the underlying material and structural behavior to the analyst than do current definitions.Copyright