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.

Consideration of Environmental Fatigue in the ASME Code for Carbon and Low-Alloy Steel Components

2003 ◽  
Author(s):  
Stan T. Rosinski ◽  
Arthur F. Deardorff ◽  
Robert E. Nickell
Keyword(s):  
Fatigue Life ◽  
Alloy Steel ◽  
Water Environment ◽  
Operating Experience ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Low Alloy Steel ◽  
Reactor Water ◽  
Asme Code ◽  
The Impact

The potential impact of reactor water environment on reducing the fatigue life of light water reactor (LWR) piping components has been an area of extensive research. While available data suggest a reduction in fatigue life when laboratory samples are tested under simulated reactor water environments, reconciliation of this data with plant operating experience, plant-specific operating conditions, and established ASME Code design processes is necessary before a conclusion can be reached regarding the need for explicit consideration of reactor water environment in component integrity evaluations. U.S. nuclear industry efforts to better understand this issue and ascertain the impact, if any, on existing ASME Code guidance have been performed through the EPRI Materials Reliability Program (MRP). Based on the MRP activities completed to date there is no need for explicit incorporation of reactor water environmental effects for carbon and low-alloy steel components in the ASME Code. This paper summarizes ongoing MRP activities and presents the technical arguments for resolution of the environmental fatigue issue for carbon and low-alloy steel locations.

Application of Corrosion Fatigue Crack Growth Rate Data to Integrity Analyses of Nuclear Reactor Vessels

1979 ◽  
Vol 101 (3) ◽  
pp. 182-190 ◽  
Author(s):  
W. H. Bamford
Keyword(s):  
Growth Rate ◽  
Fatigue Crack ◽  
Crack Growth Rate ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Water Environment ◽  
Reference Curve ◽  
Reactor Water ◽  
Reference Curves ◽  
The Impact

The methodology of fatigue crack growth analysis in evaluating structural integrity of nuclear components has been well established over the years, even to the point where a recommended practice has been incorporated in Appendix A to Section XI of the ASME Code. The present reference curve for crack growth rates of pressure vessel steels in reactor water environment was developed in 1973, and since that time a great deal of data have become available. The original curve was meant to be a bounding curve, and some recent data have exceeded it, so an important question to address is which reference curve to use for these calculations. The important features of fatigue crack growth behavior in a reactor water environment are reviewed, along with some suggested explanation for the observed environmental enhancement and overall trends. The variables which must be accounted for in any reference crack growth rate curve are delineated and various methods for accomplishing this task are discussed. Improvements to the present reference curve are suggested, and evaluated as to their accuracy relative to the present curve. The impact of the alternative curves is also evaluated through solution of an example problem. A discussion of the conservatisms included in the alternative reference curves as compared with the present reference curve is included. Also research work is identified which could lead to further improvement in the reference curves.

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.

Strain Rate Calculation Approach in Environmental Fatigue Evaluations

2010 ◽  
Author(s):  
Mark A. Gray ◽  
Matthew C. Salac ◽  
David H. Roarty ◽  
E. Lyles Cranford
Keyword(s):  
Strain Rate ◽  
Water Environment ◽  
Reactor Water ◽  
Regulatory Documents ◽  
Penalty Factor ◽  
Starting Point ◽  
Asme Code ◽  
Rate Calculation ◽  
License Renewal ◽  
Insight Into

Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.

Basis of the Upcoming Changes to the Piping Inelastic Fatigue Design Rules in the ASME Boiler and Pressure Vessel Code Section III, Division 1, Article NB-3600

2015 ◽  
Author(s):  
Timothy M. Adams
Keyword(s):  
Pressure Vessel ◽  
Service Level ◽  
Fatigue Analysis ◽  
Elastic Plastic ◽  
Fatigue Design ◽  
Division 1 ◽  
Class 1 ◽  
Asme Code ◽  
Code Changes ◽  
Near Future

In conducting a Class 1 piping analysis per the simplified rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Article NB-3600, a fatigue analysis is required per paragraph NB-3653 for both Service Level A and Service Level B. The fatigue analysis provides two options. The options are dependent on Equation 10 of subparagraph NB-3653.1. If this equation is met for a given load set pair under consideration, then the analysis proceeds directly to subparagraphs NB-3653.2 through NB-3653.5. If however, Equation 10 is exceeded, the Code allows the use of a simplified Elastic Plastic Analysis as delineated in subparagraph NB-3653.6. The first requirement of NB-3653.6 is that both Equation 12 and Equation 13 must be met. The changes in the seismic design in the last 25+ years have not been appropriately reflected in the subparagraph NB-3653.6(b) Equation 13. Also, the Code provides no clear guidance on seismic anchor motions in paragraph NB-3650. In 2012 ASME Code Committees undertook an action to address these issues. This paper provides the background and basis for Code changes that are anticipated will be implemented in the near future in paragraph NB-3653.6 of the ASME Boiler and Pressure Vessel Code, Section III, Division 1 that will address both of these issues. This implementation will make the Elastic Plastic Fatigue rules of NB-3653.6 consistent with the design by analysis approach of NB-3228.5.

Modifications of the 2016 Edition of the RCC-M Code to Account for Environmentally Assisted Fatigue

2016 ◽  
Author(s):  
Stéphan Courtin ◽  
Thomas Métais ◽  
Manuela Triay ◽  
Eric Meister ◽  
Stéphane Marie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Steering Committee ◽  
Nuclear Industry ◽  
Life Extension ◽  
Asme Code ◽  
Nickel Base ◽  
Fatigue Evaluation ◽  
International Experts

The French nuclear industry has to face nowadays a series of challenges it did not have to face a decade ago. The most significant one is to ensure a reliable and safe operation of Nuclear Power Plants (NPP) in a context of both an ageing reactor fleet and new builds. The new constructions need rules that integrate a strong operation feedback while the older NPPs need rules that will guarantee the life extension beyond 40 years of operation. In this context, a new edition of the French RCC-M Code is planned for 2016. This new edition integrates the modifications made to the Code as a result of Requests for Modification (RM), which can be submitted by anyone and which help to continuously improve the quality and robustness of the Code. Concerning fatigue analyses, the RCC-M Code steering committee has acknowledged end of 2014 the reception of two RM to modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as to integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. The contents of these two RM were based on the proposals presented in Reference [1]. AFCEN required a technical review of these two RM and this task was performed by a working group composed by French and international experts. This process concluded to the approval of these two RM to be integrated to the 2016 edition of the RCC-M Code. This paper offers a presentation of these two new Rules in Probation Phase (RPP), this format being quite similar to Code Cases proposed by ASME Code.

Environmental Fatigue Evaluation of a Korean Nuclear Power Plant

2008 ◽  
Author(s):  
Chang-Kyun Oh ◽  
Hyun-Su Kim ◽  
Hag-Ki Youm ◽  
Tae-Eun Jin ◽  
Young-Jin Kim
Keyword(s):  
Fatigue Life ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Moment Stress ◽  
Wide Range ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Reactor Pressure

In accordance with the recommendation of USNRC and the U.S. license renewal experiences, the effects of reactor coolant environment on the fatigue life have to be considered for the continued operation of operating nuclear power plants as well as for the design of new plants. Although various evaluation methodologies have been suggested to date, a wide range of comparison of the existing methodologies has not been performed. The purpose of this paper is to evaluate the environmental effects on the fatigue life of a reactor pressure vessel and ASME Class 1 piping by the various methodologies and to investigate the effects of the pressure and moment stress histories on the environmental fatigue evaluation. The evaluation results show that the environmental fatigue evaluation results based on the design cumulative usage factors for the reactor pressure vessel and ASME Class 1 piping satisfy the requirement of the ASME code except for charging nozzle. However, when using operating cumulative usage factor, the environmental fatigue evaluation result for the charging nozzle satisfies the ASME code allowable. And the effects of the pressure and moment stress histories on the environmental fatigue evaluation are considered to be small when using the modified rate approach.

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.

Strain Rate Calculation Approach in Environmental Fatigue Evaluations

2014 ◽  
Vol 136 (4) ◽  
Author(s):  
Mark A. Gray ◽  
Matthew C. Salac ◽  
David H. Roarty ◽  
E. Lyles Cranford
Keyword(s):  
Strain Rate ◽  
Water Environment ◽  
Reactor Water ◽  
Regulatory Documents ◽  
Penalty Factor ◽  
Starting Point ◽  
Asme Code ◽  
Rate Calculation ◽  
License Renewal ◽  
Insight Into

Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach (MRA) to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.

scholarly journals Consistent Creep and Rupture Properties for Creep-Fatigue Evaluation

1979 ◽  
Vol 101 (4) ◽  
pp. 276-285 ◽  
Author(s):  
C. C. Schultz
Keyword(s):  
Fatigue Life ◽  
Factor Of Safety ◽  
304 Stainless Steel ◽  
Code Design ◽  
Class 1 ◽  
Creep Fatigue ◽  
Asme Code ◽  
Accepted Practice ◽  
Fatigue Evaluation ◽  
Rupture Properties

The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a minimum strength heat is also shown to provide reasonable predictions. The viability of using consistent properties (either actual or those of a minimum strength heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations.

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