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

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

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 (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. 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.

The Impact of an Improved Flaw Model on a Pressurized Thermal Shock Evaluation

2002 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Computational Models ◽  
The United States ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Evaluation Program ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
The Impact ◽  
Flaw Characterization

The current regulations for pressurized thermal shock (PTS) were derived from computational models that were developed in the early-mid 1980s. The computational models utilized in the 1980s conservatively postulated that all fabrication flaws in reactor pressure vessels (RPVs) were inner-surface breaking flaws. It was recognized at that time that flaw-related data had the greatest level of uncertainty of the inputs required for the probabilistic-based PTS evaluations. To reduce this uncertainty, the United States Nuclear Regulatory Commission (USNRC) has in the past few years supported research at Pacific Northwest National Laboratory (PNNL) to perform extensive nondestructive and destructive examination of actual RPV materials. Such measurements have been used to characterize the number, size, and location of flaws in various types of welds and the base metal used to fabricate RPVs. The USNRC initiated a comprehensive project in 1999 to re-evaluate the current PTS regulations. The objective of the PTS Re-evaluation program has been to incorporate advancements and refinements in relevant technologies (associated with the physics of PTS events) that have been developed since the current regulations were derived. There have been significant improvements in the computational models for thermal hydraulics, probabilistic risk assessment (PRA), human reliability analysis (HRA), materials embrittlement effects on fracture toughness, and fracture mechanics methodology. However, the single largest advancement has been the development of a technical basis for the characterization of fabrication-induced flaws. The USNRC PTS-Revaluation program is ongoing and is expected to be completed in 2002. As part of the PTS Re-evaluation program, the updated risk-informed computational methodology as implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, including the improved PNNL flaw characterization, was recently applied to a domestic commercial pressurized water reactor (PWR). The objective of this paper is to apply the same updated computational methodology to the same PWR, except utilizing the 1980s flaw model, to isolate the impact of the improved PNNL flaw characterization on the PTS analysis results. For this particular PWR, the improved PNNL flaw characterization significantly reduced the frequency of RPV failure, i.e., by between one and two orders of magnitude.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

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.

An Overview of the Pressurized Thermal Shock Re-Evaluation Project

2005 ◽  
Author(s):  
Terry L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Thermal Shock ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Technical Basis ◽  
Evaluation Project

In 1999, a study sponsored by the United States Nuclear Regulatory Commission (NRC) suggested that advances in the technologies associated with the physics of pressurized-thermal-shock (PTS) events developed since the derivation of the PTS regulations (established in the early-mid eighties) had the potential to establish a technical basis that could justify a relaxation in the current PTS-related regulations. A relaxation of these regulations could have profound implications for plant license extension considerations. Subsequently, the NRC initiated the interdisciplinary PTS Re-evaluation Project. During the five year project, an updated comprehensive computational methodology evolved, within the framework established by modern probabilistic risk assessment (PRA) techniques, through interactions among experts in relevant disciplines from the NRC staff, their contractors, and representatives from the nuclear industry. During 2004, the updated computational methodology was applied to three domestic commercial pressurized water reactors (PWRs). The most recent results of the PTS Re-evaluation Project provide a technical basis to support a relaxation of the current PTS regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. The details of the updated computational methodology, the mathematical models, the analysis results, key findings, and supporting information have recently been drafted in several very detailed and lengthy formal reports. These reports are currently under review at the NRC. An objective of this paper is to provide a short overview of the improved computational methodology, analysis results, and key findings of the PTS re-evaluation project. To demonstrate that a technical basis has been established to support a relaxation of the current PTS regulations, it is helpful to understand the derivation of the current PTS regulations; therefore, another objective of this paper is to contrast the interpretation of the analysis results of the PTS re-evaluation to those performed in the eighties from which the current PTS regulations were derived.

Risk-Based Fracture Evaluation of Reactor Vessels Subjected to Cool-Down Transients Associated With Shutdown: An Examination of the Effects of Different Modeling Approaches on Estimated Failure Probabilities

2006 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Asme Code ◽  
Computational Methodology ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Evaluation Project

The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned 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 ASME Code. In 1999, the USNRC initiated the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project to determine if a technical basis could be established to support a relaxation in the current PTS regulations. The PTS re-evaluation project included the development and application of an updated risk-based computational methodology that incorporates several advancements applicable to modeling the physics of vessel fracture due to thermal hydraulic transients imposed on the RPV inner surface. The results of the PTS re-evaluation project demonstrated that there is a sound technical basis to support a relaxation of the current PTS regulations. The results of the PTS re-evaluation are currently under review by the USNRC. Based on the promising results of the PTS re-evaluation, the USNRC has recently applied the updated computational methodology to fracture evaluations of RPVs subjected to planned cool-down transients, associated with reactor shutdown, derived in accordance with ASME Section XI – Appendix G. The objective of these analyses is to determine if a sound technical basis can be established to provide a relaxation to the current regulations for the derivation of bounding cool-down transients as specified in Appendix G to Section XI of the ASME Code. This paper provides a brief overview of these analyses, results, and the implications of the results.

Alternate Requirements for Protection Against Pressurized Thermal Shock (PTS)

2013 ◽  
Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Xavier Pitoiset
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Prediction Techniques ◽  
The U.S

Plants in the United States (U.S.) and many plants outside of the U.S. are required to meet the regulations of the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61. The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61). The Alternate PTS Rule was developed through probabilistic fracture mechanics (PFM) evaluations performed for selected U.S. pilot plants. A Generalization Study was also performed which determined that the plants used for these evaluations were representative of and applicable to the U.S. Pressurized Water Reactor (PWR) nuclear power plant fleet. Plants outside of the U.S. may be interested in implementing the Alternate PTS Rule. However, direct implementation of the Alternate PTS Rule may not be possible due to differences in plant design, embrittlement prediction techniques, inservice inspection requirements, etc. The objective of this paper is to explore the use the Alternate PTS Rule by PWR plants outside of the U.S. by proposing methods to account for the potential differences mentioned above.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

The Sensitivity of Pressurized Thermal Shock Analysis Results to Alternative Models for Weld Flaw Distributions

2003 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
Evaluation Study ◽  
The United States ◽  
Metal Plate ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Pressurized Thermal Shock ◽  
Technical Basis ◽  
The Impact

The United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project in 1999 to determine if improved technologies can provide a technical basis to reduce the conservatism in the current regulations for pressurized thermal shock (PTS) 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 renewal considerations. During the PTS re-evaluation study, an improved risk-informed computational methodology was developed that provides a more realistic characterization of PTS risk. This updated methodology was recently applied to three commercial PWRs. The results of this study provide encouragement that a technical basis can be established to support a relaxation of current PTS regulations. One significant model improvement applied in the PTS re-evaluation study was the development of flaw databases derived from the non-destructive and destructive examinations of material from cancelled reactor pressure vessels (RPV). Empirically-based statistical distributions derived from these databases and expert illicitation were used to postulate the number, size, and location of flaws in welded and base metal (plate and forging) regions of an RPV during probabilistic fracture mechanics (PFM) analyses of RPVs subjected to transient loading conditions such as PTS. However, limitations in the available flaw data have required assumptions to be made to complete the risk-based flaw models. Sensitivity analyses were performed to evaluate the impact of four flaw-related assumptions. Analyses addressed: 1) truncations of distributions to exclude flaws of extreme depth dimensions, 2) vessel-to-vessel differences in flaw data, 3) large flaws observed in weld repair regions, and 4) the basis for estimating the number of surface breaking flaws. None of the four alternate weld flaw models significantly impacted calculated vessel failure frequencies or invalidated the tentative conclusions derived from the USNRC PTS re-evaluation study.

Background to Recent Revision of the Section III Seismic Piping Rules

2002 ◽  
Author(s):  
John Minichiello ◽  
Ernest B. Branch ◽  
Timothy M. Adams ◽  
Yasuhide Asada ◽  
Richard W. Barnes
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Experimental Program ◽  
Japanese Industry ◽  
Regulatory Commission ◽  
Special Working ◽  
Special Working Group ◽  
Considerable Discussion

The new rules for seismic piping design in Section III that were developed and included in the requirements in 1994 Addenda of the ASME Boiler and Pressure Vessel Code (B&PV Code) generated considerable discussion within the industry and from the United States Nuclear Regulatory Commission, (USNRC). The USNRC initiated a review of the results of the previous EPRI/NRC experimental program and the Japanese industry started its own experimental program. To accommodate and address developments resulting from these efforts, the ASME, B&PV Code established a Special Working Group (SWG) to continue the review and study of the questions and information generated. This paper reports on the efforts of this SWG which resulted in refinements of the revised rules. These refinements have been accepted for inclusion in Section III of the ASME, B&PV Code.

The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients

2008 ◽  
Author(s):  
Terry Dickson ◽  
Mark EricksonKirk
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Regulatory Commission ◽  
Technical Basis

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 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 ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also 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). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.

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