The Experts below are selected from a list of 234 Experts worldwide ranked by ideXlab platform
J. B. Spiesman - One of the best experts on this subject based on the ideXlab platform.
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Comparisons of ANS, ASME, AWS, and NFPA Standards cited in the NRC Standard Review Plan, NUREG-0800, and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANS, ASME, AWS and NFPA Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review
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Comparisons of ANSI Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANSI Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
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Comparisons of ASTM Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ASTM Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
David Rudland - One of the best experts on this subject based on the ideXlab platform.
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Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part II — Examples
Volume 6B: Materials and Fabrication, 2016Co-Authors: Robert E. Kurth, Cedric Jean-marie Sallaberry, Frederick W. Brust, E. Kurth, Michael L. Benson, David RudlandAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR) [1], defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations. Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. A companion paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. This paper illustrate with examples the implementation of the described methods into xLPR v2.0.
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Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part I — Theory
Volume 6B: Materials and Fabrication, 2016Co-Authors: Robert E. Kurth, Cedric Jean-marie Sallaberry, Frederick W. Brust, Michael L. Benson, Elizabeth A. Kurth, David RudlandAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations [1]. Using the lessons learned from the pilot study [2] the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB [3]. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential sources of uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. This paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. The current WRS sampling scheme employs correlation in order to smooth the shape of the WRS fields through the thickness of a dissimilar metal weld. This method presents an enrichment of the Cholesky decomposition on the correlation matrix, in order to satisfy the other two requirements.
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Development of the Extremely Low Probability of Rupture (xLPR) Version 2.0 Code
Volume 6B: Materials and Fabrication, 2015Co-Authors: David Rudland, Craig Harrington, Rémi DingrevilleAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations. As reported in previous technical papers, the first version of the xLPR code was developed as part of a pilot study, which leveraged existing fracture mechanics based models and software coupled to both a commercial and open source code framework to determine the framework and architecture requirements appropriate for building a modular-based code with this complexity. Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focuses on those primary piping systems previously approved for LBB. In this version, the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. This paper examines the xLPR Version 2.0 model by presenting a brief overview of the xLPR scope, the code structure, computational framework and fracture mechanics-based models. As a demonstration of the xLPR Version 2.0 capabilities, an example is presented that focuses on PWSCC in large-bore piping systems. This example exercises some functionalities of the xLPR code to demonstrate its application to assess compliance with 10CFR50 Appendix A, GDC-4. This paper concludes with a brief discussion on the path forward and Plans for the control and maintenance of the xLPR code.
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15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems‐Water Reactors - Development of the Extremely Low Probability of Rupture (xLPR) Code
15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2012Co-Authors: David Rudland, Craig HarringtonAbstract:10 CFR 50 Appendix A General Design Criteria (GDC) 4 requires that primary piping systems exhibit an extremely low probability of rupture in order to exclude dynamic effects associated with postulated primary pipe ruptures. The Leak-Before-Break (LBB) methodology, as described in NRC Standard Review Plan (SRP) 3.6.3, was developed to meet this goal. Per SRP 3.6.3, active degradation mechanisms are not permitted in systems approved for LBB and Pressurized Water Reactors (PWRs) are currently experiencing Primary Water Stress Corrosion Cracking (PWSCC). For the long term, NRC began a cooperative research program with the Electric Power Research Institute to develop a probabilistic assessment tool that quantitatively assesses the probability of primary piping system rupture. This paper provides an overview of the xLPR program, focusing on the cooperative structure used for model development and results from the proof-of-concept pilot study.
K. L. Bohlander - One of the best experts on this subject based on the ideXlab platform.
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Codes and Standards and other guidance cited in regulatory documents
1996Co-Authors: J. R. Nickolaus, K. L. BohlanderAbstract:As part of the U.S. Nuclear Regulatory Commission (NRC) Standard Review Plan Update and Development Program (SRP-UDP), Pacific Northwest National Laboratory developed a listing of industry consensus codes and Standards and other government and industry guidance referred to in regulatory documents. The SRP-UDP has been completed and the SRP-Maintenance Program (SRP-MP) is now maintaining this listing. Besides updating previous information, Revision 3 adds approximately 80 citations. This listing identifies the version of the code or Standard cited in the regulatory document, the regulatory document, and the current version of the code or Standard. It also provides a summary characterization of the nature of the citation. This listing was developed from electronic searches of the Code of Federal Regulations and the NRC`s Bulletins, Information Notices, Circulars, Enforcement Manual, Generic Letters, Inspection Manual, Policy Statements, Regulatory Guides, Standard Technical Specifications and the Standard Review Plan (NUREG-0800).
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Comparisons of ANS, ASME, AWS, and NFPA Standards cited in the NRC Standard Review Plan, NUREG-0800, and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANS, ASME, AWS and NFPA Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review
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Comparisons of ANSI Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANSI Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
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Comparisons of ASTM Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ASTM Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
A. R. Ankrum - One of the best experts on this subject based on the ideXlab platform.
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Comparisons of ANS, ASME, AWS, and NFPA Standards cited in the NRC Standard Review Plan, NUREG-0800, and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANS, ASME, AWS and NFPA Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review
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Comparisons of ANSI Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ANSI Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
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Comparisons of ASTM Standards cited in the NRC Standard Review Plan, NUREG-0800 and related documents
1995Co-Authors: A. R. Ankrum, K. L. Bohlander, E. R. Gilbert, R. A. Pawlowski, J. B. SpiesmanAbstract:This report provides the results of comparisons of the cited and latest versions of ASTM Standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC`s Standard Review Plan Update and Development Program. Significant changes to the Standards, from the cited version to the latest version, are described and discussed in a tabular format for each Standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the Standard. The results and recommendations presented in this document have not been subjected to NRC staff Review.
Frederick W. Brust - One of the best experts on this subject based on the ideXlab platform.
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Benchmarking Probabilistic Code for LBB Analysis for Circumferential Cracks
Volume 1B: Codes and Standards, 2017Co-Authors: Robert E. Kurth, Cedric Jean-marie Sallaberry, Bruce A. Young, Paul Scott, Frederick W. Brust, E. KurthAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. There are several codes available for addressing the requirements of GDC-4. This paper addresses three of these codes: (1) xLPR 2.0; (2) PROLOCA; and (3) PROMETHEUS. Each of these codes is described and applied to a representative Plant where active degradation mechanisms have been found. Conclusions about the design, results, and interpretation of the results is then provided. In all cases the probability of failure of the pipe is found to be extremely low when the crack inspections and leak detection systems are modeled.
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Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part II — Examples
Volume 6B: Materials and Fabrication, 2016Co-Authors: Robert E. Kurth, Cedric Jean-marie Sallaberry, Frederick W. Brust, E. Kurth, Michael L. Benson, David RudlandAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR) [1], defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations. Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. A companion paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. This paper illustrate with examples the implementation of the described methods into xLPR v2.0.
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Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part I — Theory
Volume 6B: Materials and Fabrication, 2016Co-Authors: Robert E. Kurth, Cedric Jean-marie Sallaberry, Frederick W. Brust, Michael L. Benson, Elizabeth A. Kurth, David RudlandAbstract:NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations [1]. Using the lessons learned from the pilot study [2] the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB [3]. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential sources of uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. This paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. The current WRS sampling scheme employs correlation in order to smooth the shape of the WRS fields through the thickness of a dissimilar metal weld. This method presents an enrichment of the Cholesky decomposition on the correlation matrix, in order to satisfy the other two requirements.
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Leak-Before-Break Under Beyond Design Basis Seismic Loading
Journal of Pressure Vessel Technology, 2013Co-Authors: Tao Zhang, Gery Wilkowski, Frederick W. Brust, Alfredo A. Betervide, Oscar MazzantiniAbstract:The Atucha II nuclear power Plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original Plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The Plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in Standard Review Plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., “Robust LBB Analysis for Atucha II Nuclear Plant,” 2011 ASME PVP Conference, July 17–21, Baltimore, MD) on the seismic evaluations for the Atucha II Plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1 × 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1 × 10−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.
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LBB Under Beyond Design Basis Seismic Loading
Volume 8: Seismic Engineering, 2012Co-Authors: Tao Zhang, Gery Wilkowski, Frederick W. Brust, Alfredo A. Betervide, Oscar MazzantiniAbstract:The Atucha II nuclear power Plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original Plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The Plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II Plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.