Issues Regarding Use of Master Curve Data in Resetting Initial RTNDT

2006 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
National Laboratory ◽  
Safety Evaluation ◽  
Nuclear Regulatory Commission ◽  
Crack Size ◽  
Oak Ridge ◽  
The Us ◽  
Large Populations ◽  
Asme Code ◽  
Shock Experiment ◽  
Regulatory Commission

The B&W Owners Group submitted justification for resetting the initial RTNDT for the Linde 80 weld materials using ASME Code Case N-629/N-631 to the US Nuclear Regulatory Commission and received the NRC Safety Evaluation Report with some adjustments. Two major issues were encountered during the review and approval process: 1) The pressurized thermal shock experiment data from Oak Ridge National Laboratory with full-length axial (very long) cracks fall below the Code Case curve. This observation led to the question whether there are implicit crack size limitations in the code case, and 2) For the large populations of data examined, a larger portion of data falls below the Code Case N-629/N-631 curve than the ASME KIC curve, prompting a question whether the code case is functionally equivalent to the ASME KIC curve. This paper describes these major issues and how they were addressed.

Analysis of Circumferential Welds in BWRs for Life Beyond 60

2015 ◽  
Author(s):  
Patrick Purtscher ◽  
Simon Sheng ◽  
Terry Dickson
Keyword(s):  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Boiling Water Reactors ◽  
Circumferential Welds ◽  
Oak Ridge ◽  
Original Analysis ◽  
Asme Code ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes the probabilistic fracture mechanics (PFM) analyzes of the conditional probability of failure (CPF) due to brittle fracture of circumferential welds (CW) from a cold overpressurize event in boiling water reactors (BWR) operated for 72 EFPY. This analysis used the Fracture Analysis for Vessels, Oak Ridge (FAVOR) computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding. Two typical vessel configurations and the associated material properties for the beltline materials, CW, axial welds (AW), and plates (PL) were used. The analyses consider the potential effects of different fabrication options, shop vs field. Shop-fabrication is mainly by submerged arc weld (SAW) process, while field fabrication used the shielded metal arc weld (SMAW) process. In either case, repairs would have required the SMAW process. The calculations show that field-fabricated vessels would have a slight increase in the CPF compared to shop-fabricated vessels, but the assumed fraction of repair welds was more significant than the fabrication option. The details demonstrate the relative importance of surface-breaking flaws vs. embedded flaws for the assumed transient. The results confirm the conclusions from the original analysis from BWRVIP-05 and BWRVIP-74, the CPF for CW is orders of magnitude less than that of PL and AW regions of the vessel; therefore, the ASME Code-required volumetric examinations of the CW every 10 years as part of the in-service inspection (ISI) program does not change the overall CPF for the vessel. In all the cases analyzed, the total CPF values of the BWRs for 72 EFPY are below the goal for safe operation.

scholarly journals Review of Practice for Deeply Embedded/Buried NPP Structures Subject to Seismic Loadings

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
National Laboratory ◽  
Earth Pressure ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Nuclear Containment ◽  
The Us ◽  
Reactor Building ◽  
Seismic Loadings ◽  
Knowledge And Practice ◽  
Regulatory Commission

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1

2010 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Computational Study ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Problem Class ◽  
Wide Range ◽  
Significant Generalization ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.

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

2011 ◽  
Author(s):  
Terry Dickson ◽  
Mark Kirk ◽  
Eric Focht
Keyword(s):  
Structural Integrity ◽  
National Laboratory ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Operating Life ◽  
Oak Ridge ◽  
Asme Code

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.

Technical Basis for Proposed Fourth Revision to ASME Code Case N-513

2014 ◽  
Author(s):  
Robert O. McGill ◽  
Guy DeBoo ◽  
Russell C. Cipolla ◽  
Eric J. Houston
Keyword(s):  
Nuclear Regulatory Commission ◽  
Short Term ◽  
Safety Issues ◽  
Evaluation Procedures ◽  
The Us ◽  
Bent Pipe ◽  
Asme Code ◽  
Moderate Energy ◽  
Regulatory Commission ◽  
Technical Basis

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been three revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the three revisions have been previously published. There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fourth revision. These changes include addressing the current condition on the Code Case acceptance by the US Nuclear Regulatory Commission (NRC), clarification of the Code Case applicability limits and expansion of Code Case scope to additional piping components. New flaw evaluation procedures are given for through-wall flaws in elbows, bent pipe, reducers, expanders and branch tees. These procedures evaluate flaws in the piping components as if in straight pipe by adjusting hoop and axial stresses to account for the geometry differences. These changes and their technical bases are described in this paper.

scholarly journals Deployment and Operation of the ES-3100 Type B Shipping Container

2006 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Dean R. Tousley ◽  
Dennis B. Miller
Keyword(s):  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Oak Ridge ◽  
Shipping Container ◽  
Enriched Uranium ◽  
Transportation Research ◽  
Regulatory Commission ◽  
The U.S

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping, for disposition purposes, bulk quantities of fissile materials, primarily highly enriched uranium (HEU). The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material since the 1980s. However, the 6M does not conform to the packaging requirements in the Code of Federal Regulations (10 CFR 71) and, for that reason, is being phased out for use in the DOE secure transportation system by the end of 2006. BWXT Y-12 developed and licensed the ES-3100 container to replace the DOT 6M. The ES-3100 was certified by the Nuclear Regulatory Commission (NRC) in April 2006. The process of deploying the new package began in June 2005 and is planned to be completed in July 2006. The package will be fully operational and completely replace the DOT 6M at the Y-12 National Security Complex (Y-12) by October 2006. This paper reviews the deployment process and the mock loading station that was installed at National Transportation Research Center (NTRC) of Oak Ridge National Laboratory. Specialized equipment, tools, and instrumentation that support the handling and loading operations of the ES-3100 are described in detail. Loading options for other user sites are explored in preparation for deployment of this new state-of-the-art shipping container throughout the DOE complex and the private sector.

Response of a Dry Storage Spent Fuel Cask to Hypothetical Loading

2005 ◽  
Author(s):  
Gustavo A. Aramayo ◽  
Douglas J. Ammerman ◽  
Jeffrey A. Smith
Keyword(s):  
Analytical Methods ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Spent Fuel ◽  
Oak Ridge ◽  
Transportation Risk ◽  
Dry Storage ◽  
Regulatory Commission ◽  
The U.S

This paper addresses the analytical methods used to determine the response of a dry storage spent fuel cask to hypothetical loading. Because of the sensitive nature of the topic under discussion, the response of the cask is described in qualitative terms, and the paper is intentionally vague on the parameters and results. This research was sponsored by the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office. The work was performed under contract from the Sandia National Laboratory (SNL), Transportation Risk and Packing organization. The analytical effort was performed at the Oak Ridge National Laboratory (ORNL) facilities with loading specified by SNL.

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