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 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.

An Improvement in the Code Evaluation of Flawed Pipes

2007 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Base Material ◽  
Limit Load ◽  
Evaluation Methodology ◽  
Linear Elastic ◽  
Weld Material ◽  
Steel Base ◽  
Asme Code ◽  
Elastic Fracture

ASME Code Section XI Nonmandatory Appendix C [1] formalized evaluation of flaws in piping for justification of continued service of piping components with an identified crack-like flaw. The revision of this appendix in 2004 was a significant improvement in the evaluation methodology for both flawed austenitic stainless steel and ferritic steel pipe depending upon the failure mode governed by limit load (fully plastic), elastic-plastic fracture mechanics, or linear elastic fracture mechanics. The appendix also provides a screening procedure to determine failure mechanism and a procedure for flaw modeling based on the estimated flaw size at the end of a specified evaluation period. The purpose of this paper is to propose an improvement to the limit load method applicable to screened-in carbon steel, wrought stainless steel base material, stainless steel weld material with nonflux weld, and cast products in which the ferrite content is less than twenty percent. In addition, changes in the formulation are proposed to extend the methodology to non-crack-like flaws. Both crack-like and non-crack-like circumferential flaws in the piping are analyzed to simplify formulation for flaw evaluation. The paper concludes that the proposed formulation improves efficiency of the application of Appendix C methodology for crack-like flaw and non-crack-like flaw evaluations.

Flaw Proximity Rules for Parallel Planar Flaws Under Limit Load Conditions

2005 ◽  
Author(s):  
Sam Ranganath ◽  
Bob Carter ◽  
Francis Ku ◽  
Marcos Herrera
Keyword(s):  
Fracture Mechanics ◽  
Test Data ◽  
Limit Load ◽  
Linear Elastic ◽  
Parallel Cracks ◽  
Asme Code ◽  
Elastic Fracture ◽  
Load Analysis ◽  
Load Conditions ◽  
Reactor Internals

This paper describes the evaluation of adjacent cracks (including parallel cracks in different planes) from a limit load perspective. The proposed approach complements the current ASME Code rules that are largely based on linear elastic fracture mechanics considerations. The evaluation considers limit load analysis and is validated by comparison with test data. In particular the paper provides criteria to determine load capability of structures with offset cracks in parallel planes. The objective of the work was to provide criteria for the evaluation of reactor internals, but the results can be applied to evaluate cracking in ductile pressure boundary piping also.

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.

Development of EPFM Procedure for Axially Flawed Pipe Using Z Factor Based on CVN

2006 ◽  
Author(s):  
Kunio Hasegawa ◽  
Katsumasa Miyazaki ◽  
Naoki Miura ◽  
Koich Kashima ◽  
Douglas A. Scarth
Keyword(s):  
Fracture Mechanics ◽  
Ductile Fracture ◽  
Working Group ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Empirical Equations ◽  
Shelf Energy ◽  
Evaluation Procedures ◽  
Asme Code ◽  
Semi Empirical

Evaluation procedures on an allowable axial flaw in a pipe for fully plastic fracture is provided by limit load criteria in Appendix C-5000 of the ASME Code Section XI. However, flaw evaluation for ductile fracture using EPFM (Elastic Plastic Fracture Mechanics) criteria is not provided for axial flaw in the Appendix. Methodology of the flaw evaluation for ductile fracture using EPFM criteria is discussing at the Working Group on Pipe Flaw Evaluation of ASME Code Section XI. Many failure experiments on axially flawed pressurized pipes made of moderate toughness materials had been performed at Battelle Columbus Laboratories. Semi-empirical equations for predicting failure stresses were developed from these experiments. This paper describes a derivation of load multiplier, Z factor, based on Charpy V notch upper shelf energy (CVN) from failure stresses for moderate toughness materials based on the experiments, and proposes a flaw evaluation procedure to determine allowable axial flaw for a ductile fractured pipe using the EPFM criteria.

scholarly journals Application of the FAVOR-OCI Fracture Mechanics Computer Program to ASME Code Section XI, IWB-3610 Flaw Acceptance Criteria Evaluations

2018 ◽  
Author(s):  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Critical Values ◽  
Initiation Toughness ◽  
Acceptance Criteria ◽  
Oak Ridge ◽  
The Us ◽  
Asme Code

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR. The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws. The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Reactor Vessel Nozzle Inner Radius Fracture Analyses Using Elastic-Plastic Fracture Mechanics

2018 ◽  
Author(s):  
S. E. Marlette
Keyword(s):  
Fracture Mechanics ◽  
The United States ◽  
Reactor Vessel ◽  
Nondestructive Examination ◽  
Elastic Plastic ◽  
Linear Elastic ◽  
Inner Radius ◽  
Asme Code ◽  
American Society ◽  
Ductile Behavior

The American Society of Mechanical Engineers (ASME) published Section XI Code Case N-648-1 [1] in order to provide alternative examinations of reactor vessel nozzle inner radii. The Code Case was created because ultrasonic examination of the inner radius regions of reactor vessels nozzles is not practical within the operating fleet and the likelihood of flaws developing within these locations is extremely low. Justification for using alternative visual examinations was provided in a paper published at the 2001 Pressure Vessel and Piping (PVP) Technology Conference [2]. This 2001 PVP paper used linear elastic fracture mechanics (LEFM) to demonstrate tolerance for flaws significantly larger than would be detected using nondestructive examination techniques. However, the Code Case [1] and PVP paper [2] were only applicable to operating plants in the United States. Thus, there was a need to provide a similar fracture analysis considering the AP1000® design to support elimination of volumetric examinations of the nozzle inner radius regions. It was also important to consider improvements in facture mechanics techniques that have been recently published in the ASME Code. The ductile behavior of the material at operating temperatures allow for the use of elastic plastic fracture mechanics (EPFM) methods which provides significantly improved flaw tolerance results. This paper compares results from analyses using LEFM and the EPFM methods for the AP1000 reactor vessel nozzle inner radii region and demonstrates tolerance for large flaws within these regions in order to support a basis for elimination of volumetric inspection during in-service and pre-service examination for the AP1000 design.

Fully Plastic Failure Stresses and Allowable Crack Sizes for Circumferentially Surface Cracked Pipes Subjected to Tensile Loading

2021 ◽  
Author(s):  
Kunio Hasegawa ◽  
David Dvorak ◽  
Vratislav Mares ◽  
Bohumir Strnadel ◽  
Yinsheng Li
Keyword(s):  
Experimental Data ◽  
Bending Moment ◽  
Limit Load ◽  
Tensile Loading ◽  
Crack Size ◽  
Plastic Failure ◽  
Circumferential Crack ◽  
Asme Code ◽  
American Society ◽  
Better Than

Abstract Fully plastic failure stresses for circumferentially surface cracked pipes subjected to tensile loading can be estimated by means of limit load criteria based on the net-section stress approach. Limit load criteria of the first type (labelled LLC-1) were derived from the balance of uniaxial forces. Limit load criteria of the second type are given in Section XI of the ASME (American Society of Mechanical Engineering) Code, and were derived from the balance of bending moment and axial force. These are labelled LLC-2. Fully plastic failure stresses estimated by using LLC-1 and LLC-2 were compared. The stresses estimated by LLC-1 are always larger than those estimated by LLC-2. From the literature survey of experimental data, failure stresses obtained by both types of LLC were compared with the experimental data. It can be stated that failure stresses calculated by LLC-1 are better than those calculated by LLC-2 for shallow cracks. On the contrary, for deep cracks, LLC-2 predictions of failure stresses are fairly close to the experimental data. Furthermore, allowable circumferential crack sizes obtained by LLC-1 were compared with the sizes given in Section XI of the ASME Code. The allowable crack sizes obtained by LLC-1 are larger than those obtained by LLC-2. It can be stated that the allowable crack size for tensile stress depends on the condition of constraint of the pipe, and the allowable cracks given in Section XI of the ASME Code are conservative.

3D Residual Stress Simulation of an Excavate and Weld Repair Mockup

2016 ◽  
Author(s):  
Francis H. Ku ◽  
Pete C. Riccardella ◽  
Steven L. McCracken
Keyword(s):  
Residual Stresses ◽  
Weld Metal ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Weld Repair ◽  
Element Analysis ◽  
Pressurized Water ◽  
Piping Systems ◽  
Asme Code ◽  
Original Construction

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.

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