ASME Section III Flaw Tolerance Sample Problem for Fatigue Design of Nuclear System Components

2014 ◽  
Author(s):  
David J. Dewees ◽  
Paul Hirschberg ◽  
Wolf Reinhardt ◽  
Gary L. Stevens ◽  
David H. Roarty ◽  
...  
Keyword(s):  
Environmental Effects ◽  
Water Environment ◽  
Sample Problem ◽  
Design Assessment ◽  
Thermal Transients ◽  
Flaw Tolerance ◽  
Initial Task ◽  
Class 1 ◽  
Fatigue Evaluation ◽  
Technical Basis

An ASME Section III Task Group (TG) was formed in 2012 to develop alternate rules for the design assessment of Section III Class 1 nuclear components subject to fatigue service with environmental effects. Specifically, a flaw tolerance approach is being investigated based on similar methodology to that found in ASME Section XI Nonmandatory Appendix L. A key initial task of the TG (which reports to the Section III Working Group on Environmental Fatigue Evaluation Methods) was to develop and solve a detailed sample problem. The intent of the sample problem was to illustrate application of proposed rules, which will be documented as a Section III Code Case with a supporting technical basis document. Insights gained from round robin solution of the sample problem are presented and discussed in this paper. The objective of documenting the findings from the sample problem are to highlight the observed benefits and limitations of the proposed procedures, particularly how rules typically associated with in-service experience might be adapted into design methods. The sample problem is based on a heavy-walled stainless steel nozzle that meets cumulative fatigue usage requirements in air (i.e., usage factor, U, without reactor water environment effects less than unity), but fails to meet usage factor requirements when environmental fatigue effects are applied. The sample problem demonstrates that there is a class of problems dominated by severe thermal transients where fatigue initiation is predicted based on elastic methods including environmental effects, but fatigue crack propagation results are acceptable. Preliminary conclusions are drawn based on the results of the sample problem, and the next steps are also identified.

Fatigue Evaluation for Section III, Class 1 Components Using Flaw Tolerance Method to Consider Environmental Effects

2015 ◽  
Author(s):  
David Roarty ◽  
Wolf Reinhardt ◽  
David Dewees
Keyword(s):  
Fatigue Crack ◽  
Environmental Effects ◽  
Design Assessment ◽  
Flaw Tolerance ◽  
Design Life ◽  
Tolerance Approach ◽  
Class 1 ◽  
Fatigue Evaluation ◽  
Tolerance Method ◽  
Type 304

An ASME Section III Task Group (TG) was formed in 2012 to develop alternate rules for the design assessment of Section III Class 1 nuclear components subject to fatigue service with environmental effects. A Section III Code Case has been proposed with the purpose of providing a method for performing fatigue evaluations of Class 1 components when the effects of a light water reactor environment on fatigue life are judged to be significant and cumulative usage factor (CUF) limits may not be satisfied. The Code Case implements a flaw tolerance approach by postulating that a fatigue crack initiates at the beginning of life and is subjected to fatigue crack growth under the specified design cycles. It must be demonstrated that the crack would remain stable with set margin throughout the design life of the component or part under consideration, and would remain confined to an acceptable fraction of the wall thickness. At this time, the application is limited to type 304/304L and 316/316L austenitic steel. This paper discusses the methodology and technical background of the proposed Code Case.

Environmental Fatigue Evaluation for Primary Components of APR1400 Reactor Coolant System

2010 ◽  
Author(s):  
J. M. Kim ◽  
K. W. Kim ◽  
K. S. Yoon ◽  
S. H. Park ◽  
I. Y. Kim ◽  
...  
Keyword(s):  
Environmental Effects ◽  
Integral Method ◽  
Power Reactor ◽  
Light Water Reactor ◽  
Outlet Nozzle ◽  
Design Life ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Coolant System

USNRC Regulatory Guide (RG) 1.207 provides a guideline for evaluating fatigue analyses due to the environmental effects on the new light water reactor (LWR). The environmental correction factor (Fen) is used to incorporate the LWR environmental effect into fatigue analyses of ASME Class 1 components. In this paper, the environmental fatigue evaluation is applied to some primary components with 60 year design life of Advanced Power Reactor (APR1400). The materials sampled from Class 1 components are the low alloy steel for the reactor vessel (RV) outlet nozzle and the carbon steel for the hot leg which are attached to the outlet nozzle. The simplified method, time-based integral method and strain-based integral method are used to compute the Fen values. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. As the calculated cumulative fatigue usage factor considering environmental effects (CUFen) is below 1.0, there is no concern for the RV outlet nozzle to implement design for environmental fatigue effects.

JSME Codes on Environmental Fatigue Evaluation

2006 ◽  
Author(s):  
Takao Nakamura ◽  
Makoto Higuchi ◽  
Takehiro Kusunoki ◽  
Yasuaki Sugie
Keyword(s):  
Fatigue Life ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Water Environment ◽  
Cooling Water ◽  
Practical Method ◽  
Plant Equipment ◽  
Fatigue Evaluation

The “Guidelines for Evaluating Fatigue Initiation Life Reduction in the LWR Environment” (the MITI Guidelines) including equations to evaluate environmental fatigue were issued and notified the electric utilities in September 2000 by the former Agency for Natural Resources and Energy in Japan. The MITI Guidelines require the Japanese utilities to take into account environmental effects when conducting fatigue evaluation associated with Plant Life Management (PLM) activities for operating nuclear power plants. However, the MITI guidelines do not specify how to conduct the environmental fatigue evaluation under actual plant conditions. To provide a concrete and practical method to deal with environmental effects on fatigue evaluation of plant equipment, Thermal and Nuclear Power Engineering Society established the “Guidelines on Environmental Fatigue Evaluation for LWR Component” (the TENPES Guidelines) in 2002. Since then, the Japan Society of Mechanical Engineers (JSME) has reviewed the equations to calculate the environmental fatigue life correction factor, Fen in the MITI guidelines and the methods to evaluate the environmental fatigue in the TENPES guidelines considering the latest environmental fatigue data. Based on the result of the review, JSME intends to establish new environmental fatigue evaluation method. This paper explains the scheme and the technical basis of the evaluation methods in JSME codes, and the positioning of the codes to apply them to actual plant conditions. Another paper is released separately that shows the background of the equation to evaluate the fatigue life under the reactor cooling water environment.[19]

Status Report on ASME Section III Subgroup on Design Plan for Code Changes to Implement Environmental Fatigue Evaluation Methods

2010 ◽  
Author(s):  
Jack R. Cole ◽  
John C. Minichiello
Keyword(s):  
Action Plan ◽  
Water Environment ◽  
Nuclear Industry ◽  
Status Report ◽  
Reactor Water ◽  
Class 1 ◽  
Asme Code ◽  
Code Changes ◽  
Fatigue Evaluation ◽  
The Impact

This paper provides a status report on the ASME Section III Subgroup on Design Environmental Fatigue Action Plan. The plan will direct development of ASME Section III Code [1] changes to provide guidance on acceptable methods for evaluating reactor water environment effects on reactor coolant pressure boundary components. Section III provides indication to the user that special consideration should be given for the environment to which a component is exposed, but does not provide guidance in addressing these effects. Discussions on needed ASME Code changes to address reactor water environmental effects have been under consideration by ASME Code bodies for many years. Due to the renaissance of the nuclear industry it is now apparent that Section III should be up-dated to address the missing guidance. The action plan was developed by the Subgroup on Design to coordinate activities necessary for Code bodies to act on proposed Code changes that will provide the user with the necessary tools to evaluate the effect of reactor water environment on fatigue life of components. The action plan lays out a strategy for a staged implementation of analysis methodologies, needed research, analysis guides, sample problems, and an assessment of the impact of the new rules upon the industry. The ultimate goal of the Subgroup on Design is to develop a new non-mandatory appendix that provides guidance to the user when evaluating reactor water environmental fatigue effects on Class 1 components.

Technical Basis of Proposed New Acceptance Standards for Class 1, 2 and 3 Piping

2007 ◽  
Author(s):  
Katsumasa Miyazaki ◽  
Kunio Hasegawa ◽  
Naoki Miura ◽  
Koichi Kashima ◽  
Douglas A. Scarth
Keyword(s):  
Fracture Mechanics ◽  
Ultrasonic Inspection ◽  
Limit Load ◽  
Flaw Size ◽  
High Fracture Toughness ◽  
Evaluation Procedure ◽  
High Fracture ◽  
Development Methodology ◽  
Class 1 ◽  
Technical Basis

Acceptance Standards in Section XI of the ASME Boiler and Pressure Vessel Code have an important role as the first step in the flaw evaluation procedure. When a flaw size is within the allowable flaw size in the Acceptance Standard, the flaw is acceptable and analytical evaluation is not required. Although ASME Section XI has Acceptance Standards for Class 1 piping in IWB-3500, there are no Acceptance Standards for Class 2 and 3 piping. Furthermore, the development of the current Acceptance Standards for Class 1 piping was based on flaw detectability by ultrasonic inspection and consideration of fracture mechanics. In this paper, the development of proposed new Acceptance Standards for Class 2 and 3 piping, as well as for Class 1 piping, is described. The development methodology is based on a fracture mechanics approach. For Class 1 piping with high fracture toughness, the allowable flaw sizes were determined by limit load solution. For Class 1 piping, the intent was to maintain overall consistency with the current Acceptance Standards. Proposed Acceptance Standards for Class 2 and 3 austenitic piping were also developed by the methodology used to develop the proposed new Acceptance Standards for Class 1 piping. Allowable flaw sizes for both surface flaws and subsurface flaws for preservice and inservice examinations were developed.

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

2021 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Thermal Transients ◽  
Piping Systems ◽  
Class 1 ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Comparison of International Codes for a Fatigue Crack Growth Flaw Tolerance Sample Problem

2021 ◽  
Author(s):  
Gary L. Stevens
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Large Diameter ◽  
Sample Problem ◽  
Pressurized Water ◽  
Typical Application ◽  
Flaw Tolerance ◽  
Asme Code ◽  
The Impact

Abstract As part of the development of American Society of Mechanical Engineers Code Case N-809 [1], a series of sample calculations were performed to gain experience in using the Code Case methods and to determine the impact on a typical application. Specifically, the application of N-809 in a fatigue crack growth analysis was evaluated for a large diameter austenitic pipe in a pressurized water reactor coolant system main loop using the current analytical evaluation procedures in Appendix C of Section XI of the ASME Code [2]. The same example problem was previously used to evaluate the reference fatigue crack growth curves during the development of N-809, as well as to compare N-809 methods to similar methods adopted by the Japan Society of Mechanical Engineers. The previous example problem used to evaluate N-809 during its development was embellished and has been used to evaluate additional proposed ASME Code changes. For example, the Electric Power Research Institute investigated possible improvements to ASME Code, Section XI, Nonmandatory Appendix L [3], and the previous N-809 example problem formed the basis for flaw tolerance calculations to evaluate those proposed improvements [4]. In addition, the ASME Code Section XI, Working Group on Flaw Evaluation Reference Curves continues to evaluate additional research data and related improvements to N-809 and other fatigue crack growth rate methods. As a part of these Code investigations, EPRI performed calculations for the Appendix L flaw tolerance sample problem using three international codes and standards to evaluate fatigue crack growth (da/dN) curves for PWR environments: (1) ASME Code Case N-809, (2) JSME Code methods [5], and (3) the French RSE-M method [6]. The results of these comparative calculations are presented and discussed in this paper.

Guidelines on Environmental Fatigue Evaluation for LWR Component

2003 ◽  
Author(s):  
Takao Nakamura ◽  
Itaru Saito ◽  
Yasuhide Asada
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Reduction Factor ◽  
Engineering Society ◽  
Fatigue Initiation ◽  
Plant Life ◽  
Actual Operating ◽  
Fatigue Evaluation ◽  
Fatigue Life Reduction ◽  
Operating Plant

Japanese utilities and vendors have taken environmental effects on fatigue (EF) into consideration in the plant life management (PLM) activity of operating plants for several years. In Sep. 2000 MITI notified the utilities to adopt “The Guidelines for Evaluating Fatigue Initiation Life Reduction in LWR Environment (MITI guidelines)” for PLM evaluation of operating plants [1]. In April 2001, the study started to establish detailed procedures for EF evaluation and the committee was organized for developing detailed guidelines at Thermal and Nuclear Power Engineering Society (TENPES). The evaluation guidelines were completed and published as TENPES guidelines [2]. These guidelines proposed several practical options to apply fatigue life reduction factor for environmental effects (Fen) on actual operating plant fatigue evaluation.

Development of Instrumentation and Control Systems for UK ABWR

2017 ◽  
Author(s):  
Itsuki Naito ◽  
Taisuke Koyamada ◽  
Keisuke Yamamoto ◽  
Kingo Igarashi ◽  
Hideo Harada ◽  
...  
Keyword(s):  
Control System ◽  
Control Systems ◽  
Good Practice ◽  
Operating Experience ◽  
Severe Accident ◽  
Regulatory Regime ◽  
Design Assessment ◽  
Class 1 ◽  
The Uk ◽  
And Control

This paper introduces the Instrumentation and Control (I&C) system for the proposed UK Advanced Boiling Water Reactor (UK ABWR) offered by Hitachi-GE Nuclear Energy, Ltd (Hitachi-GE). Hitachi-GE has been progressing the UK Generic Design Assessment (GDA) licensing process over the last 3 years. This is the process through which the Office for Nuclear Regulations (ONR) assesses the UK ABWR for suitability from a nuclear safety, security, environmental protection and waste management perspective and it is the first step towards proceeding with the construction phase in the UK. ONR’s regulatory expectations setting out relevant good practice are described in the Safety Assessment principles (SAPs), which are considered into the I&C design for UK ABWR. In addition, it has also been designed to take into account relevant good practices and regulations. In accordance with expectations derived from SAPs, the UK ABWR I&C systems are categorized and classified as required by IEC 61513 and IEC 61226. In addition, the overall I&C architecture, including all associated Human-Machine Interfaces (HMIs), abides by the principles independence and diversity of safety measures, segregation and separation of the protection and control systems. As a result, the UK ABWR I&C architecture is composed of major eight sub-systems. The eight sub-systems are: -Safety System Logic and Control system (SSLC) -Hardwired Backup System (HWBS) -Safety Auxiliary Control System (SACS) -Plant Control System (PCntlS) -Reactor/Turbine Auxiliary Control System (RTACS) -Plant Computer System (PCS) -Severe Accident Control and Instrumentation system (SA C&I) -Other dedicated C&I systems. The features for each sub-system such as redundancy of safety train or segregation among divisions are specified so that each sub-system will achieve its reliability as well as increase availability. While in the Japanese ABWR safety I&C system, the main protection system (SSLC), is microprocessor-based from the decades of successful operating experience in the past BWR, to meet the UK regulatory regime expectation on diversity between Class 1 platform and non-Class 1 platform, the SSLC (Class 1) for the UK ABWR is by Field Programmable Gate Array (FPGA). This system is currently under development and complies with IEC 62556. Its safety integrity level is planned to be SIL 3 (as a single division) and SIL 4 (as a four division system) as defined in IEC 61508. The HMIs which constitute an integral part of the I&C systems are also designed to comply with the I&C architecture regarding their categorization and classification with consideration of Human Factors (HF) modern methods taken into accounts.

Comparison of Strain Rate Calculation Methods for Environmental Fatigue Correction Evaluation

2012 ◽  
Author(s):  
Seiji Asada
Keyword(s):  
Strain Rate ◽  
Principal Stress ◽  
Evaluation Method ◽  
Calculation Methods ◽  
Sample Problem ◽  
Nuclear Facility ◽  
Principal Stress Direction ◽  
Stress Direction ◽  
Rate Calculation ◽  
Fatigue Evaluation

A Code Case for procedure to determine strain rate and Fen for environmental fatigue evaluation is under preparation in the ASME BPV Committee on Construction of Nuclear Facility Components (III). The draft Code Case is to incorporate two methods for strain rate calculation. One is based on NB-3216.1 “Constant Principal Stress Direction” that comes from the JSME Environmental Fatigue Evaluation Method. The other is based on NB-3216.2 “Varying Principal Stress Direction” that was proposed by M. Gray et al. In this paper, both methods are explained and compared by using a sample problem.

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