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.

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.

Flaw Evaluation and Allowable Flaw Sizes of Nuclear Piping in JSME Code

2003 ◽  
Author(s):  
Kunio Hasegawa ◽  
Hideo Kobayashi ◽  
Koichi Kashima
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Japan Society ◽  
Acceptance Criteria ◽  
Evaluation Code ◽  
Evaluation Procedures ◽  
Class 1 ◽  
American Society ◽  
Service Inspection

A flaw evaluation code for nuclear power plants has been developed at the Japan Society of Mechanical Engineers (JSME) in 2000 and revised adding inspection rules in 2002. Then the code consists of inspection for nuclear components and evaluation procedures of flaws in Class 1 components detected during in-service inspection. This paper introduces the summary of the JSME Code and describes two kinds of allowable flaw sizes, Acceptance Standards and Acceptance Criteria, for Class 1 pipes in the flaw evaluation procedures. In addition, these allowable flaws are compared with those in the ASME (American Society of Mechanical Engineers) Code Section XI.

Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

2013 ◽  
Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Factors ◽  
Design Code ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Current Design ◽  
Factor Design ◽  
Excessive Load ◽  
Load And Resistance Factor

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

What’s New in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components

2004 ◽  
Author(s):  
Gary Park
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Mechanical Engineering ◽  
Nuclear Industry ◽  
Nondestructive Examination ◽  
American Society ◽  
Plant Components

The nuclear industry is a pretty dynamic industry, in that it is always on the move, changing every time we turn around. For that very reason, there is a need to keep up with the industry by providing changes to American Society of Mechanical Engineering Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” There have been many changes over the last three years. This paper addresses a few of those, but gives a feel for the number of changes from the 2000 Addenda to the 2003 Addenda, there have been a total of approximately 56 changes. Of those changes, 11 were in the repair/replacement requirements, 19 in the inspection requirements, 4 in the evaluation requirements, 18 in the nondestructive examination requirements, and 4 in the administrative requirements. The paper classifies the changes as “Technically Significant,” “Significant,” “Non-Significant,” or “Editorial.” The paper addresses only a few of those changes that were “Technically Significant.” The paper also includes some of the activities that the ASME Section XI Subcommittee is currently working on.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

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.

Comparison of Allowable Bending Moments for Circumferentially Flawed Pipes in ASME Boiler and Pressure Vessel Code Section XI Nonmandatory Appendices C and H

2019 ◽  
Author(s):  
Hune-Tae Kim ◽  
Ji-Su Kim ◽  
Jun-Min Seo ◽  
Yun-Jae Kim ◽  
Kuk-Hee Lee ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Bending Moments ◽  
Analytical Evaluation ◽  
Evaluation Procedures ◽  
Failure Assessment ◽  
Inner Surface ◽  
Plant Components

Abstract In this paper, allowable bending moments for circumferential inner surface cracked pipes are evaluated. ASME Boiler and Pressure Vessel code Section XI, “Rules for Inspection of Nuclear Power Plant Components” provides analytical evaluation procedures. Analytical evaluation methods based on the failure mechanism are provided in nonmandatory Appendix C and those based on failure assessment diagram are given in nonmandatory Appendix H. Allowable bending moments are evaluated using both appendices and compared with experiments. Conservativeness is compared quantitatively between both methods by normalizing allowable bending moments with experimental maximum moments.

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.

BWR Steam Dryer Issues and Lessons Learned Related to Flow Induced Vibration

2013 ◽  
Author(s):  
Chakrapani Basavaraju ◽  
Kamal A. Manoly ◽  
Martin C. Murphy ◽  
William T. Jessup
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Thermal Power ◽  
Lessons Learned ◽  
Flow Induced Vibration ◽  
Safety Function ◽  
Water Reactors ◽  
American Society ◽  
Loose Parts

Steam dryers in Boiling Water Reactors, located in the upper steam dome of the reactor pressure vessel, are not pressure retaining components and are not designed and constructed to the provisions of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. As such, these components do not correspond to any specific safety class referenced in the Code. Although the steam dryers in BWRs perform no safety function, they must maintain the structural integrity in order to avoid the generation of loose parts that may adversely impact the capability of other plant equipment to perform their safety functions. Therefore guidance from Section III of the ASME Code is utilized in the design and fabrication of replacement dryers as well as for design modifications of the existing dryers for extended power uprates. The majority of licensees of operating nuclear plants are applying for EPU, which generally increases the thermal power output to 20% above the original licensed thermal power. Nuclear power plant components such as steam dryers can be subjected to strong fluctuating loads and can experience unexpected high cycle fatigue due to adverse flow effects while operating at EPU conditions. However, there are some unique challenges related to steam dryer operation at EPU conditions requiring special considerations to prevent fatigue damage from the effects of flow induced vibration. This paper examines the issues and lessons learned related to FIV considerations during EPU reviews of BWR steam dryers.

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