Collapse of Nuclear Reactor SG Tubes Pressurized From Outside: The Influence of Imperfections

2010 ◽  
Vol 133 (1) ◽  
Author(s):  
Lelio Luzzi ◽  
Valentino Di Marcello
Keyword(s):  
Numerical Investigation ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
External Pressure ◽  
Initial Stresses ◽  
Asme Code ◽  
Conservative Attitude ◽  
Computer Codes ◽  
American Society ◽  
Collapse Behavior

Some innovative nuclear power plant proposals consider for the design tubes of considerable thickness subjected to external pressure (e.g., steam generators tubes). The collapse of thick tubes is expected to be dominated by yielding but, because of the decreasing nature of the postcollapse evolution, interaction with buckling is likely to be significant enough to demand consideration. At the present, few studies have been carried out both experimentally and numerically, as witnessed by the really conservative attitude that codes assume for thick tubes. A numerical investigation has been performed in this context at the Politecnico di Milano, which was originally intended as a support for requesting a relaxation of American Society of Mechanical Engineers (ASME) regulations. Actually, in 2007, ASME code case N-759 was approved, permitting significant thickness saving in the tube design. Nevertheless, the numerical investigation was pursued to assess the influence of different parameters, such as eccentricity, initial stresses, and material hardening, on the collapse of tubes with diameter to thickness ratios D/t<20. Results are thought to be useful under at least two respects: first, providing some understanding on the collapse behavior in a thickness range so far unexplored; second, giving an indication on the assumptions on which computer codes ought to be based when numerical analyses are required.

Influence of Imperfections on the Collapse of Thick Tubes Subjected to External Pressure

2009 ◽  
Author(s):  
Valentino Di Marcello
Keyword(s):  
Numerical Investigation ◽  
Nuclear Reactors ◽  
Oil Industry ◽  
External Pressure ◽  
Initial Stresses ◽  
Primary System ◽  
Steam Generators ◽  
Water Pipelines ◽  
Conservative Attitude ◽  
Computer Codes

Cylindrical shells pressurized from outside are required for several engineering applications, and a growing need of tubes with significant thickness has been recently experienced in the oil industry (very deep water pipelines) and in the frame of integrated primary system nuclear reactors (steam generators). Their collapse behaviour has been explored little if at all, both experimentally and numerically, as witnessed by the extremely conservative attitude that codes assume for very thick tubes. A numerical investigation has been performed in this context at the Politecnico di Milano, which was originally intended as a support for requesting a relaxation of ASME regulations. In fact, in 2007 Code Case N-759 [1] was approved, permitting significant thickness saving in the tube design. Nevertheless, the numerical investigation was pursued to assess the influence of different parameters, such as eccentricity, initial stresses and material hardening, on the collapse of tubes with diameter to thickness ratios D/t&lt;20. Results are thought to be useful under at least two respects: first, they provide some understanding on the collapse behaviour in a thickness range so far unexplored; secondly, they give an indication on the assumptions on which computer codes ought to be based when numerical analyses are required.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

2014 ◽  
Author(s):  
Jason Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Diagnostic Testing ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Light Water ◽  
Water Reactor ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Components in ASME Code Section XI

2003 ◽  
Author(s):  
Russell C. Cipolla ◽  
Guy H. DeBoo ◽  
Warren H. Bamford ◽  
Kenneth K. Yoon ◽  
Kunio K. Hasegawa
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Primary Objective ◽  
Surveillance Program ◽  
Acceptance Criteria ◽  
Evaluation Procedures ◽  
Asme Code ◽  
Nuclear Component ◽  
American Society ◽  
Plant Components

The primary objective of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI is to provide the rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Pressure boundary integrity in terms of assuring resistance to sudden and catastrophic failure has been an essential objective of the ASME Code since its inception in 1914. These objectives are especially important in ASME Section XI since maintaining pressure boundary integrity of components has a crucial role in ensuring safe and reliable operation in nuclear operating plants. The purpose of this paper is to describe the evaluation procedures, methods, and acceptance criteria for flaws detected in plant components during implementation of in-service inspection surveillance program. For nuclear plant components, pressure boundary integrity includes both leak integrity (no leakage from the reactor coolant system) and structural integrity (no rupture or burst of the pressure boundary). The evaluation requirements in ASME Section XI provide specific rules for assessing the acceptance limits for flaw indications that may be detected during the service life of a nuclear component. In addition to describing current flaw evaluation procedures, details of recent Code developments and improvements are discussed.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Piping in ASME Code Section XI

2003 ◽  
Author(s):  
Douglas A. Scarth ◽  
Gery M. Wilkowski ◽  
Russell C. Cipolla ◽  
Sushil K. Daftuar ◽  
Koichi K. Kashima
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Limit Load ◽  
Acceptance Criteria ◽  
Linear Elastic ◽  
Evaluation Procedures ◽  
Piping Systems ◽  
Asme Code ◽  
American Society ◽  
Plant Evaluation

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provides rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Evaluation procedures and acceptance criteria for the evaluation of flaws in nuclear piping in Section XI of the ASME Code were first published in 1983 and have been under revision for the past several years. This paper provides an overview of the procedures and acceptance criteria for pipe flaw evaluation in Section XI. Both planar and nonplanar flaws are addressed by Section XI. The evaluation procedures and acceptance criteria cover: failure by plastic collapse as characterized by limit load analysis; fracture due to ductile tearing prior to attainment of limit load, as characterized by elastic-plastic fracture mechanics (EPFM) analysis; and brittle fracture as characterized by linear elastic fracture mechanics (LEFM) analysis. A major revision to the evaluation procedures and acceptance criteria was published in the 2002 Addenda to Section XI. Evaluation procedures and acceptance criteria in the 2001 Edition, as well as the revisions in the 2002 Addenda, are described in this paper. Code Cases that address evaluation of wall thinning in piping systems, as well as temporary acceptance of flaws in moderate energy piping systems, are also described.

Comparison of Fatigue Crack Growth Curves of Japan, the United States, and European Union Code and Standards

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Kiminobu Hojo ◽  
Yukio Takahashi
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Nuclear Power ◽  
Growth Curves ◽  
The United States ◽  
Austenitic Steels ◽  
Engineering Society ◽  
Asme Code ◽  
American Society

There are several codes, standards, handbooks, and guidelines for the nuclear power plant maintenance in Japan, the US, and EU. They include Stress Corrosion Cracking (SCC) and fatigue crack growth curves for crack growth calculation. In this paper, the authors selected five kinds of codes, standards and guidelines, and compared their fatigue crack growth curves for choice of the suitable curves. The feature of each curve was quantitatively evaluated. Japan Society of Mechanical Engineers (JSME) maintenance rule and American Society of Mechanical Engineers (ASME) code provide the fatigue crack growth formulae for both ferritic and austenitic steels and consider the environmental effects in some cases. The Fitness-for-Service Network (FITNET) curves are categorized in many kinds of metal, whereas the Forschungskuratorium Maschinenbau (Germany) = Board of Trustees of Mechanical Engineering (FKM) guideline and Welding Engineering Society (WES) procedure provide the common properties generally applicable to steels.

Validation of Pipe Flaw Evaluation Procedures in Section XI of the ASME Code Against Pipe Fracture Experiments

2004 ◽  
Author(s):  
Phuong H. Hoang ◽  
Gery M. Wilkowski
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Limit Load ◽  
Acceptance Criteria ◽  
Linear Elastic ◽  
Evaluation Procedures ◽  
Asme Code ◽  
Elastic Fracture ◽  
American Society ◽  
Plant Evaluation

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provides rules and requirements for maintaining pressure boundary integrity of piping during the life of a nuclear power plant. Evaluation procedures and acceptance criteria for the evaluation of flaws in nuclear piping in Section XI of the ASME Code were first published in 1983 and have been under revision for the past several years. The evaluation procedures and acceptance criteria cover: failure by plastic collapse as characterized by limit load analysis; fracture due to ductile tearing prior to attainment of limit load, as characterized by elastic-plastic fracture mechanics (EPFM) analysis; and brittle fracture as characterized by linear elastic fracture mechanics (LEFM) analysis. A major revision to the evaluation procedures and acceptance criteria was published in the 2002 Addenda to Section XI. A brief overview of the pipe flaw evaluation procedures published in the 2002 Addenda are provided in the paper. The evaluation procedures that were published in the 2002 Addenda have been validated against the results of a large number of pipe fracture experiments. The results of this validation exercise are summarized in this paper.

Solution of a Sample Problem Related to Revision 1 of Code Case N-830

2017 ◽  
Author(s):  
Mark Kirk ◽  
Steven Xu ◽  
Cheng Lui ◽  
Marjorie Erickson ◽  
Yil Kim ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Temperature Dependence ◽  
Master Curve ◽  
Nuclear Reactor ◽  
Sample Problem ◽  
Asme Code ◽  
American Society ◽  
Technical Basis ◽  
Direct Use ◽  
Current Code

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case (CC) N-830. CC N-830 permits the direct use of fracture toughness in flaw evaluations as an alternative to the indirect/correlative approaches (RTNDT-based) traditionally used in the ASME Code. The current version of N-830 estimates allowable fracture toughness values in the transition regime as the 5th percentile Master Curve (MC) indexed to the transition temperature T0. The proposed CC N-830 revision expands on this capability by incorporating a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J–R) used currently, or useful in, a flaw evaluation for conditions ranging from the lower shelf through the upper shelf. Papers presented in previous ASME Pressure Vessel and Piping (PVP) Conferences since 2014 provide the technical basis for these various toughness models. This paper contributes to this overall CC N-830 documentation suite by presenting the results of a sample problem run to assess the proposed revision of the CC. The objective of the sample problem was (1) to determine if the revised CC was written with adequate clarity to permit different engineers to accurately and consistently calculate the various allowable toughness values described by the equations in the CC, (2) to assess how these allowable toughness values would be used to calculate allowable flaw depths using standard ASME SC-XI approaches, and (3) to compare allowable flaw depths calculated using established Code practices (RTNDT-based) to those calculated using proposed CC practices (T0-based). The sample problem demonstrated that (1) the CC was written with sufficient clarity to allow different engineers to arrive at the same estimated value of allowable toughness, (2) the latitude associated with the provisions of the ASME Code pertinent to estimation of allowable flaw depth are responsible for some differences in the allowable flaw depth values reported by different participants, and (3) current Code estimates of allowable flaw depth are far more conservative (that is: smaller) than values estimated by the candidate CC methods based on the MC, this mostly due to the generally-conservative bias of the Code’s RTNDT & KIc approach. The candidate CC methods provide much more consistent conservatism than current Code approaches for all conditions in the operating nuclear reactor fleet via their use of an index temperature (T0) defined by actual fracture toughness data and a temperature dependence defined by those data. The WGFE is continuing to evaluate candidate approaches to estimate allowable toughness values for CC N-830 using a T0-indexed Master Curve. Associated work is addressed by two companion papers presented at this conference.

Reconstitution of Reactor Pressure Boundary Components Design Stress Report for License Renewal

2010 ◽  
Author(s):  
Kirby Woods ◽  
Kenneth Thomas
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Cyclic Life ◽  
System Supplier ◽  
Asme Code ◽  
American Society ◽  
Renewal Period ◽  
License Renewal ◽  
Design Stress

The majority of United States Commercial Nuclear Power Plants (CNPP) within the next 10 years will reach the end of their license to operate and can be renewed per the “Atomic Energy Act” of 1954. This act allowed the commission to issue commercial electric power nuclear plants a license to operate (“licensed, but not to exceed 40 years, and maybe renewed upon the expiration of such period, (Chapter 10, Sec. 103(c))).” These CNPP licenses are also governed by the NSSS vendor specification requirements and by the American Society of Mechanical Engineers (ASME) design code standards. This connection is in the form of a stress report that defines the “cyclic life” adequacy criteria for this operational limit of 40 years. The license extension subsequently requires a reconstitution of the initial design stress report input parameters per ASME IWA-4120, 4223 & 4311 (e) for the renewal period extension. This requirement can entail an analysis of the operating conditions and cycles to demonstrate the material elasticity is maintained. The proposed approach for this reconstitution effort was a reanalysis in the form of a study of the Nuclear Steam System Supplier (NSSS) vendors’ original methodology to determine the NSSS vendor specification requirement for ASME code compliance and “cyclic life” adequacy. The information acquired from this evaluation has demonstrated the application to be a complex and simplistic approach. This effort to unravel the composite loading (thermal and pressure transients) condition in relation to specific plant and incident cycles provides both an understanding of component end-of-life limits and supports a comprehensive template for future fatigue life management programs. This paper summarizes this reconstituting effort that utilizes the original vessel stress analysis report to support the license renewal effort, provides a template for future fatigue management programs, demonstrates the conservatism of design, and offers an insight into the design philosophy revealing an elegant process that assures against failures.

scholarly journals ASME OM Code Implementation Lessons Learned for the AP1000® Plant

2017 ◽  
Author(s):  
Augustine A. Cardillo ◽  
Natalie M. Rodgers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Federal Regulations ◽  
Plant Design ◽  
Regulatory Requirements ◽  
Implementation Challenges ◽  
Service Testing ◽  
Asme Code ◽  
American Society

The Title 10 of the Code of Federal Regulations (10 CFR) Part 52 process and unique aspects of a passive plant design have presented new challenges for the development and implementation of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements. This paper will discuss lessons learned from the development and implementation of pre-service testing (PST) / in-service testing (IST) program plans for the AP1000® plant, for both international and domestic. Topics to be addressed include the following: • Level of detail in design certification • Treatment of unique passive plant features • Design certification commitments beyond Code requirements • Future regulatory requirements for high-risk non-safety component PST/IST • Implementation challenges for international plants Paper published with permission.

ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Robert B. Keating ◽  
Suzanne P. McKillop ◽  
Todd Allen ◽  
Mark Anderson
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Department Of Energy ◽  
Research Project ◽  
Compact Heat Exchangers ◽  
High Temperature Reactor ◽  
American Society

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHXs) for the next generation of nuclear reactors. The DOE is sponsoring research to support the development and deployment of CHXs for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) that includes university research institutes, national laboratories, manufacturers, and industry experts. The objective is to enable the use of CHX designs in advanced reactor service. A necessary step for achieving this objective is to ensure that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a CHX in an advanced reactor. A strategy for ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will also be required. The purpose of this ASME Code Roadmap is to identify the research gaps impeding the development of suitable construction and ISI rules for CHXs for high temperature reactor service and to provide a framework to utilize the research project results consistent with the expectations and needs of the industry and future owners.

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