3D Linear Bifurcation Analysis of Steel Containment Vessel

2016 ◽  
Author(s):  
Jay Schmidt ◽  
Bernd Laskewitz
Keyword(s):  
Bifurcation Analysis ◽  
Nuclear Regulatory Commission ◽  
Asymmetric Structure ◽  
Plant Design ◽  
Containment Vessel ◽  
Asme Code ◽  
Containment Shell ◽  
Axisymmetric Structure ◽  
The Stability ◽  
Regulatory Commission

The AP1000® Containment Vessel (CV) is a freestanding steel containment designed to protect the public from radiation release. The CV consists of 2 ellipsoidal heads connected by a cylindrical shell and is constructed of carbon steel. The AP1000 plant design has four large penetrations (two airlocks and two equipment hatches) located in approximately the same quadrant of the circumference of the shell which imposes asymmetric effects in the shell structure. The CV is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Subsection NE. Traditionally, the local and global stability of freestanding steel containments have been designed by use of formulae using conservative assumptions based on an axisymmetric structure. ASME Code Case N-284 “Metal Containment Shell Buckling Design Methods, Class MC Section III, Division 1” outlines methodology for satisfying the stability of the CV using two approaches. Section 1710 provides a stress based buckling approach using detailed formulae that assumes an axisymmetric structure. The second approach provides guidance and acceptability based on a linear bifurcation analysis (2D (1720) or 3D (1730)). Due to the asymmetric structure of the CV, the 3D linear bifurcation method delivers the most accurate results. The methodology and assumptions implemented by Westinghouse to qualify stability of the CV via Code Case N-284 are outlined. Also, the procedure to properly amplify the stresses as required by N-284 is included as justification of the methods used. This justification was thoroughly investigated by the Nuclear Regulatory Commission (NRC) and deemed acceptable.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

Field Experience With Code Case N-659: Ultrasonic Examination in Lieu of Radiography

2007 ◽  
Author(s):  
Douglas O. Henry
Keyword(s):  
Nuclear Power ◽  
Cost Benefit ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Material Transport ◽  
Ultrasonic Examination ◽  
Key Factor ◽  
Asme Code ◽  
Regulatory Commission ◽  
Sample Set

Code Case N-659 Revision 0 was approved in 2002 to allow ultrasonic examination (UT) an alternative to radiography (RT) for nuclear power plant components and transport containers under Section III of the ASME Code. The Nuclear Regulatory Commission has not approved N-659 and its subsequent revisions (currently N-659-2) for general use, but they have been used on a case-by-case basis mainly where logistic problems or component configuration have prevented the use of radiography. Like the parallel Code Case 2235 for non-nuclear applications under Section I and Section VIII, Code Case N-659 requires automated, computerized ultrasonic systems and capability demonstration on a flawed sample as a prerequisite for using UT in lieu of RT. Automated ultrasonic examination can be significantly more expensive than radiography, so a cost-benefit evaluation is a key factor in the decision to use the Code Case. In addition, the flaw sample set has recently become an issue and a topic of negotiation with the NRC for application of the Case. A flaw sample set for a recent radioactive material transport cask fabrication project was successfully negotiated with the NRC. The Code Case N-659 approach has been used effectively to overcome barriers to Code required radiography. Examples are examination of welds in an assembled heat exchanger and in a radioactive material transport cask assembly where internal shielding prevented radiography of the weld. Future development of Code Case N-659 will address sample set considerations and application-specific Code Cases, such as for storage and transport containers, will be developed where NRC concerns have been fully addressed and regulatory approval can be obtained on a generic basis.

Non-Proprietary Application of ASME Code Section III, Appendix N, to SONGS Replacement Steam Generators

2017 ◽  
Author(s):  
R. D. Blevins
Keyword(s):  
Elastic Stability ◽  
Nuclear Regulatory Commission ◽  
Stability Diagram ◽  
Design Flow ◽  
Mode Shapes ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Public Data ◽  
Asme Code ◽  
Regulatory Commission

Flow-induced vibration analysis of the San Onofre Nuclear Generating Station (SONGS) Replacement Steam Generators is made using non-proprietary public data for these steam generators on the Nuclear Regulatory Commission public web site, www.NRC.com. The analysis uses the methodology of Appendix N Section III of the ASME Boiler and Pressure Vessel Code, Subarticle N-1300 Flow-Induced Vibration of Tubes and Tube Banks. First the tube geometry is assembled and overall flow and performance parameters are developed at 100% design flow, then analysis is made to determine the flow velocity in the gap between tubes and tube natural frequencies and mode shapes. Finally, the mass damping and reduced velocity for tubes on the U bend are assembled and plotted on the ASME code Figure N-11331-4 fluid elastic stability diagram.

Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials

2005 ◽  
Author(s):  
V. N. Shah ◽  
B. Shelton ◽  
R. Fabian ◽  
S. W. Tam ◽  
Y. Y. Liu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Fissile Material ◽  
Radioactive Materials ◽  
Asme Code ◽  
Material Transportation ◽  
Accident Conditions ◽  
Regulatory Commission ◽  
High Level

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.

LBB Under Beyond Design Basis Seismic Loading

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Heavy Water Reactor ◽  
Regulatory Commission

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.

Developments on ASME Code Cases to Risk-Inform Repair/Replacement Activities in Support of Risk-Informed Regulation Initiatives

2002 ◽  
Author(s):  
Kenneth R. Balkey ◽  
William C. Holston
Keyword(s):  
Nuclear Energy ◽  
Nuclear Regulatory Commission ◽  
Special Treatment ◽  
Federal Regulations ◽  
The Past ◽  
Implementation Guideline ◽  
Asme Code ◽  
Informed Treatment ◽  
Regulatory Commission ◽  
Initial Results

ASME Code Case N-658, “Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities” and Code Case N-660, “Alternative Repair/Replacement Requirements For Items Classified In Accordance With Risk-Informed Processes” are being completed to expand the breadth of risk-informed requirements for pressure-retaining items. This initiative, which is built from prior ASME Section XI risk-informed inservice inspection developments over the past decade, has been undertaken in conjunction with U.S. risk-informed regulation efforts. The U.S. Nuclear Regulatory Commission (NRC) is working with the industry on risk informing Title 10 Code of Federal Regulations Part 50 (10CFR50). The Nuclear Regulatory Commission’s basic proposal is to allow modification of some of the special treatment requirements of 10CFR50. Their effort is proceeding via an Advanced Notice of Public Rulemaking, March 3, 2000, and an announcement of Availability of Draft Rule Wording, November 29, 2001, to add 10 CFR 50.69, “Risk-Informed Treatment of Structures, Systems and Components.” A parallel task by the Nuclear Energy Institute (NEI) to develop a guideline on how to implement the results of the rulemaking is also well underway via NEI 00-04 (Draft Revision B), “Option 2 Implementation Guideline,” May 2001. This paper summarizes the content and status of approval of the proposed ASME Code Cases, including how they relate to the above NRC and NEI efforts. Some initial results from trial application of the Code Cases will also be cited.

ASME Code Needs for Very High Temperature Generation IV Reactors

2009 ◽  
Author(s):  
William J. O’Donnell ◽  
Donald S. Griffin
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Operating Conditions ◽  
Asme Code ◽  
Section Viii ◽  
Construction Permit ◽  
Regulatory Commission ◽  
Very High

Subsection NH “Components in Elevated Temperature Service” of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code. The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. This paper (1)identifies the structural integrity issues in the ASME Boiler and Pressure Vessel Code, including Section II, Section VIII, Section III, Subsection NH (Class 1 Components in Elevated Temperature Service) and Code Cases that must be resolved in order to support licensing of Generation IV Nuclear Reactors, particularly Very High Temperature Gas-Cooled Reactors (VHTRs); (2) describes how the Code addresses these issues; and (3) identifies the needs for additional criteria to cover unresolved structural integrity concerns for very-high-temperature service.

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.

Implementation of ASME Code, Section XI, Code Case N-770, on Alternative Examination Requirements for Class 1 Butt Welds Fabricated With Alloy 82/182

2011 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Michael T. Anderson
Keyword(s):  
Stress Corrosion Cracking ◽  
Nuclear Regulatory Commission ◽  
Butt Welds ◽  
Division 1 ◽  
Class 1 ◽  
Final Rule ◽  
Asme Code ◽  
Areas Of Concern ◽  
Regulatory Commission ◽  
The U.S

In May 2010, the U.S. Nuclear Regulatory Commission (NRC) issued a proposed notice of rulemaking (75 FR 24324) [1] that includes a new section to its rules to require licensees to implement ASME Code Case N–770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without the Application of Listed Mitigation Activities, Section XI, Division 1,” [2] with 15 conditions. Code Case N-770 contains baseline and inservice inspection (ISI) requirements for unmitigated Alloy 82/182 butt welds and preservice and ISI requirements for mitigated Alloy 82/182 butt welds. The NRC stated that application of ASME Code Case N-770 is necessary because the inspections currently required by the ASME Code, Section XI, were not written to address stress corrosion cracking of Alloy 82/182 butt welds, and the safety consequences of inadequate inspections can be significant. The NRC expects to issue the final rule incorporating this Code Case into its regulations toward the middle of 2011. This paper discusses the new examination requirements, the conditions that NRC proposed to impose, and potential areas of concern with implementation of the new Code Case.

Impact of Undetected Fabrication Flaws on LBB Risk

2017 ◽  
Author(s):  
Robert Kurth ◽  
Cédric Sallaberry ◽  
Elizabeth Kurth ◽  
Frederick Brust
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Leak Detection ◽  
Nuclear Regulatory Commission ◽  
Acceptance Testing ◽  
Analysis Tool ◽  
Risk Of Rupture ◽  
Asme Code ◽  
Regulatory Commission ◽  
The Impact

On-going assessments of the impact of active degradation mechanisms in US nuclear power plants previously approved for leak before break (LBB) relief are being performed with, among other methods, the extremely low probability of rupture (xLPR) code being developed under a memorandum of understanding between the US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) [1]. This code finished with internal acceptance testing in July of 2016 and is undergoing sensitivity and understanding analyses and studies in preparation for its general release. One of the key components of the analysis tool is the integration of inspection methods for damage and the impact of leak detection on the risk of a pipe rupture before the pipe is detected to be leaking. While it is not impossible to detect a crack or defect that is less than 10% of the pipe wall thickness current ASME code does not give credit for inspections identifying a crack of this size. This study investigates the impact of combining the probabilistic analysis results from xLPR with a pre-existing flaw to determine the change, if any, to the risk. Fabrication flaws were found to have a statistically significant impact on the risk of rupture before leak detection.

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