Use of NUREG/CR-6909 Environmentally-Assisted Fatigue Rules for a Nuclear Plant License Renewal Application

2011 ◽  
Author(s):  
William F. Weitze ◽  
Matthew C. Walter ◽  
Keith R. Evon
Keyword(s):  
Stainless Steels ◽  
Water Environment ◽  
Austenitic Stainless Steels ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Plant ◽  
Lessons Learned ◽  
Low Alloy Steels ◽  
Alloy Steels ◽  
Regulatory Commission

As part of the process of renewing the operating license for an additional 20 years after the original 40-year design life, nuclear plant owners in the United States (US) are required to show that they are managing the effects of aging of systems, structures, and components. US Nuclear Regulatory Commission (NRC) report NUREG-1801, the “Generic Aging Lessons Learned (GALL) Report,” identifies acceptable aging management programs, including programs for fatigue and cyclic operation. This includes fatigue usage analyses that account for reduced fatigue life for components in a reactor water environment. Earlier revisions of the GALL report required plants to perform environmentally-assisted fatigue (EAF) analyses using the rules in reports NUREG/CR-6583 (for carbon and low alloy steels) and NUREG/CR-5704 (for austenitic stainless steels), which were developed in 1998 and 1999, respectively. However, GALL Revision 2, issued in December 2010, requires that the rules in NUREG/CR-6909, issued in 2007, be used for nickel alloy materials, and allows it to be used for carbon, low-alloy and stainless steels as an alternative to those in the previous reports. This paper presents an application of the NUREG/CR-6909 rules, and makes several observations about the differences between using the newer and older rules. The analyses presented were performed for a sample set of boiling water reactor (BWR) locations.

Development of New Design Fatigue Curves in Japan: Proposal of a New Fatigue Evaluation Method

2018 ◽  
Author(s):  
Seiji Asada ◽  
Akihiko Hirano ◽  
Toshiyuki Saito ◽  
Yasukazu Takada ◽  
Hideo Kobayashi
Keyword(s):  
Stainless Steels ◽  
Evaluation Method ◽  
Austenitic Stainless Steels ◽  
Fatigue Tests ◽  
Carbon Steels ◽  
Low Alloy Steels ◽  
Stress Effect ◽  
Fatigue Data ◽  
Alloy Steels ◽  
Fatigue Evaluation

In order to develop new design fatigue curves for carbon steels & low-alloy steels and austenitic stainless steels and a new design fatigue evaluation method that are rational and have clear design basis, Design Fatigue Curve (DFC) Phase 1 subcommittee and Phase 2 subcommittee were established in the Atomic Energy Research Committee in the Japan Welding Engineering Society (JWES). The study on design fatigue curves was actively performed in the subcommittees. In the subcommittees, domestic and foreign fatigue data of small test specimens in air were collected and a comprehensive fatigue database (≈6000 data) was constructed and the accurate best-fit curves of carbon steels & low-alloy steels and austenitic stainless steels were developed. Design factors were investigated. Also, a Japanese utility collaborative project performed large scale fatigue tests using austenitic stainless steel piping and low-alloy steel flat plates as well as fatigue tests using small specimens to obtain not only basic data but also fatigue data of mean stress effect, surface finish effect and size effect. Those test results were provided to the subcommittee and utilized the above studies. Based on the above studies, a new fatigue evaluation method has been developed.

Proposal of Fatigue Life Equations for Carbon and Low-Alloy Steels and Austenitic Stainless Steels as a Function of Tensile Strength

2013 ◽  
Author(s):  
Hiroshi Kanasaki ◽  
Makoto Higuchi ◽  
Seiji Asada ◽  
Munehiro Yasuda ◽  
Takehiko Sera
Keyword(s):  
Tensile Strength ◽  
Fatigue Life ◽  
Stainless Steels ◽  
Room Temperature ◽  
Austenitic Stainless Steels ◽  
Low Alloy Steels ◽  
Strength Based ◽  
Fatigue Data ◽  
Current Design ◽  
Alloy Steels

Fatigue life equations for carbon & low-alloy steels and also austenitic stainless steels are proposed as a function of their tensile strength based on large number of fatigue data tested in air at RT to high temperature. The proposed equations give a very good estimation of fatigue life for the steels of varying tensile strength. These results indicate that the current design fatigue curves may be overly conservative at the tensile strength level of 550 MPa for carbon & low-alloy steels. As for austenitic stainless steels, the proposed fatigue life equation is applicable at room temperature to 430 °C and gives more accurate prediction compared to the previously proposed equation which is not function of temperature and tensile strength.

scholarly journals Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

2008 ◽  
Vol 2008 ◽  
pp. 1-13 ◽  
Author(s):  
Robert P. Martin ◽  
Larry D. O'Dell
Keyword(s):  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Lessons Learned ◽  
Technical Program ◽  
Methodology Development ◽  
Analysis Methodology ◽  
Loss Of Coolant Accident ◽  
Core Cooling ◽  
Regulatory Commission

The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA) analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical considerations were revealed with the move to a production application. This paper describes the methodology development within the CSAU framework and utility objectives, lessons learned, and insight about current LOCA issues.

Plan and Status of Development of Design Fatigue Curves: Phase 1

2013 ◽  
Author(s):  
Seiji Asada ◽  
Akihiko Hirano ◽  
Masao Itatani ◽  
Munehiro Yasuda ◽  
Takehiko Sera ◽  
...  
Keyword(s):  
Stainless Steels ◽  
Phase 1 ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Carbon Steels ◽  
Low Alloy Steels ◽  
Energy Research ◽  
Research Committee ◽  
Engineering Society ◽  
Alloy Steels

In order to develop and propose new design fatigue curves for austenitic stainless steels, carbon steels and low alloy steels that are rational and have clear design basis, Design Fatigue Curve (DFC) subcommittee has been established in the Atomic Energy Research Committee in the Japan Welding Engineering Society and the study on design fatigue curves are going on. This paper introduces the plan and status of the activities of the DFC subcommittee.

License Renewal Demonstration Project

2002 ◽  
Author(s):  
Barry J. Elliot ◽  
Jerry Dozier
Keyword(s):  
United States ◽  
State Of The Art ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Management Program ◽  
Lessons Learned ◽  
Demonstration Project ◽  
Aging Effects ◽  
Regulatory Commission ◽  
License Renewal

Generic Aging Lessons Learned (GALL) report, License Renewal Standard Review Plan (SRP-LR), and regulatory guide were issued by the United States Nuclear Regulatory Commission (NRC) in June 2001. The intent of these documents was to provide the technical and process basis that will lead to a more effective, efficient and predictable license renewal process for industry and the NRC. The GALL report provides the aging effects on components and structures, identifies the relevant existing plant programs, and evaluates the program attributes to manage aging effects for License Renewal. The GALL report also identifies when existing plant programs would require further evaluation for License Renewal. The SRP-LR allows the applicant to reference the GALL report to demonstrate that the programs at the applicant’s facility correspond to those reviewed and approved in the GALL report. Programs that correspond to those in the GALL report will not need further detailed review by the staff. Implementation of the aging management program are verified as part of the license renewal inspection program. The GALL report identifies one acceptable way of demonstrating that components and structures have adequate aging management programs. However, applicants may propose alternatives to the programs identified in GALL. During the license renewal review, the NRC primarily focuses on areas where existing programs should be augmented or new programs developed for License Renewal. This paper will provide an overview of these documents and some of the lessons learned during a demonstration project in the application of the new guidance. This topic will be of interest to the U.S. participants considering License Renewal and desiring to know state-of-the-art information about License Renewal in the United States.

The Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials

2006 ◽  
Author(s):  
Omesh K. Chopra ◽  
William J. Shack
Keyword(s):  
Stainless Steels ◽  
Type Strain ◽  
Water Flow Rate ◽  
Austenitic Stainless Steels ◽  
Environmental Parameters ◽  
Light Water Reactor ◽  
Low Alloy Steels ◽  
Alloy Steels ◽  
Reactor Materials ◽  
American Society

The existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for carbon and low–alloy steels and austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, flow rate, surface finish, and heat-to-heat variation, on the fatigue lives of these steels are summarized. An updated version of the ANL statistical models is presented for estimating the fatigue ε–N curves for these steels as a function of the material, loading, and environmental parameters. The Fen (environmental fatigue correction factor) approach for incorporating the effects of LWR coolant environments into the fatigue evaluations of the American Society of Mechanical Engineers Code is presented.

Application of Common Basis Stress Evaluation Methodology for Environmentally Assisted Fatigue for a Pressurized Water Reactor

2013 ◽  
Author(s):  
William F. Weitze ◽  
Timothy D. Gilman ◽  
Lora Drenth
Keyword(s):  
Water Environment ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Evaluation Methodology ◽  
Pressurized Water ◽  
Stress Evaluation ◽  
Common Basis ◽  
Consistent Manner ◽  
Regulatory Commission ◽  
License Renewal

United States (US) Nuclear Regulatory Commission (NRC) report NUREG-1801, the Generic Aging Lessons Learned (GALL) Report [1], identifies acceptable aging management programs, including programs for fatigue and cyclic operation, and is used by the NRC to evaluate license renewal applications. This includes fatigue usage analyses that account for reduced fatigue life for components in a reactor water environment. Originally, it was considered acceptable for the purpose of license renewal to evaluate only the sample locations identified in NUREG/CR-6260. Recently, however, the NRC staff has been requesting license renewal applicants to demonstrate that the locations identified in NUREG/CR-6260 are the limiting locations for environmentally assisted fatigue (EAF), such that EAF evaluations are limited to the NUREG/CR-6260 locations. Any locations not bounded by the NUREG/CR-6260 locations would then be specifically addressed. A methodology has been developed for EPRI, called common basis stress evaluation (CBSE), to perform simplified stress and fatigue usage analyses in a consistent manner such that many locations can be compared for fatigue usage with and without the use of EAF, even if prior analyses of these locations were not consistent in level of detail, and even if some locations have never been evaluated for fatigue usage. This paper presents the first application of this methodology.

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