Terry L. Dickson

An updated probabilistic fracture mechanics methodology for application to pressurized thermal shock

The US Nuclear Regulatory Commission (NRC) is, in concert with the US nuclear industry, currently revisiting its rule and analysis requirements for pressurized thermal shock (PTS) scenarios. This paper provides an overview of an updated probabilistic risk analysis (PRA) methodology that is continuing to evolve as part of that effort. The evolution process includes a careful assessment of recent advancements that have been made in the various parts of the computational methodologies. The process also involves interactions between experts in relevant disciplines (thermal hydraulics, PRA, materials, fracture mechanics, and non-destructive and destructive examination). Representatives include staff members from the USNRC staff, research laboratories, and the nuclear industry. The updated methodology is being integrated into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code for application to re-examine the adequacy of the current regulations and to determine if the updates provide sufficient technical bases for revisions. This paper also discusses recent modifications to the probabilistic fracture mechanics (PFM) methodology that is central to FAVOR.

Weibull statistical models of KIc/KIa fracture toughness databases for pressure vessel steels with an application to pressurized thermal shock assessments of nuclear reactor pressure vessels

Abstract This paper presents new statistical representations of recently extended fracture toughness K Ic and K Ia databases for pressure vessel steels. These models were developed by the Heavy Section Steel Technology program at Oak Ridge National Laboratory in support of the current effort by the U.S. Nuclear Regulatory Commission to update its regulatory guidance for pressurized-thermal-shock (PTS) transients in nuclear reactor pressure vessels. The Weibull distribution, with two of its parameters calculated by the Method of Moments point-estimation technique, forms the basis for the new statistical models. An application of the new K Ic / K Ia models, as implemented in the favor probabilistic fracture mechanics computer program, is also presented for three PTS transients.

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

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.© 2003 ASME

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

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

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.Copyright

Applicability of (K, T-Stress) Methodology to Analyze RPV Under Thermal-Hydraulic Transients

The (K, T-stress) methodology developed by Gao and Dodds [1] is being utilized to introduce crack front plasticity with constraint effects when plastic deformation occurs in structures, for example, when the Reactor Pressure Vessels (RPVs) are subjected to thermal-hydraulic loadings. One crucial step in this procedure is to quantify combinations of flaw geometries and loading conditions (transient sequences) that illustrate the limits of applicability of the two-parameter (K, T-stress) advanced fracture methodology relevant to integrity analyses of RPVs subjected to normal and emergency operating conditions. Numerical analyses were conducted to determine the limits of applicability of (K, T-stress) advanced fracture technology for RPV under thermal-hydraulic loadings. The numerical results indicate that the (K, T-stress) methodology captures the constraint condition of the RPV with typical embedded flaws under a postulated dominant thermal-hydraulic transient.© 2007 ASME