Application of Draft Regulatory Guide DG-1144 Guidelines for Environmental Fatigue Evaluation to a BWR Feedwater Piping System

2007 ◽  
Author(s):  
Hardayal S. Mehta ◽  
Henry H. Hwang
Keyword(s):  
Fatigue Life ◽  
Thermal Cycle ◽  
Light Water Reactor ◽  
Piping System ◽  
Low Alloy Steel ◽  
Light Water ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Cycle Diagram ◽  
Analytical Expressions

Recently published Draft Regulatory Guide DG-1144 by the NRC provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light water reactor (LWR) environment. The analytical expressions and further details are provided in NUREG/CR-6909. In this paper, the environmental fatigue rules are applied to a BWR feedwater line. The piping material is carbon steel (SA333, Gr. 6) and the feedwater nozzle material is low alloy steel (SA508 Class 2). The transients used in the evaluation are based on the thermal cycle diagram of the piping. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. In both cases the cumulative fatigue usage factors are shown to be less than 1.0.

Evaluation of Fatigue Life of Pressurized Water Reactor Internals Considering Light-Water Reactor Coolant Environmental Effect for Aging Management Program

2014 ◽  
Author(s):  
Jianfeng Yang ◽  
Paul O’Brien
Keyword(s):  
Fatigue Life ◽  
Environmental Effect ◽  
Light Water Reactor ◽  
Pressurized Water ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Coolant ◽  
Plant Life ◽  
Fatigue Evaluation ◽  
Reactor Internals

Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.

An Investigation on the Reduction Schemes of Environmental Fatigue Usage Factor for Pressurizer Surge Line

2011 ◽  
Author(s):  
Sun-yeh Kang ◽  
Won-ho Jo ◽  
Min-sup Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Light Water Reactor ◽  
Life Extension ◽  
Regulatory Requirement ◽  
Light Water ◽  
Plant Life ◽  
Asme Code ◽  
Surge Line ◽  
Critical Components ◽  
Fatigue Evaluation

For plant life extension, it is the regulatory requirement to assess reactor coolant environmental impacts on critical components of the nuclear power plant including at least those mentioned in NUREG/CR-6260[2]. The pressurizer surge line is the most easy-to-fail component in view of LWR (Light Water Reactor) environments when it comes to meeting the current ASME code limit of the fatigue evaluation. Cumulative Usage Factor (CUF) value could be increased to a maximum of 15.35 times due to the environmental effects, which makes it easy to exceed the allowable fatigue limit (1.0). This paper discusses the process of the environmental correction factor calculation described in NUREG/CR-5704[4], and five proposed schemes for reducing the environmental CUF value to the ASME code limit or below. This paper concludes that the proposed schemes are effective in lowering the environmental CUF value of the pressurizer surge line.

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.

Sample Environmental Fatigue Calculations Associated With the Review of Draft Regulatory Guide DG-1144

2007 ◽  
Author(s):  
Gary L. Stevens ◽  
J. Michael Davis ◽  
Les Spain
Keyword(s):  
Stainless Steel ◽  
Lessons Learned ◽  
Light Water Reactor ◽  
Low Alloy Steels ◽  
Light Water ◽  
New Model ◽  
Asme Code ◽  
Alloy Steels ◽  
Reactor Materials ◽  
License Renewal

Draft Regulatory Guide DG-1144 “Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors”, July 2006 [1], and Associated Basis Draft Document NUREG/CR-6909 (ANL-06/08), “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials”, July 2006 [6] provided methods for addressing environmentally assisted fatigue (EAF) in all new nuclear plant designs. In these documents, a new model was proposed that more accurately accounts for actual plant conditions. The new model includes an EAF correction factor, Fen, which is different from Fen methods previously and currently being considered for adoption into the ASME Code. The Fen methods proposed in DG-1144 are also different than the Fen methods utilized by license renewal applicants, as required by the Generic Aging Lessons Learned (GALL) report [2], as documented in NUREG/CR-5704 [4] (for stainless steel) and NUREG/CR-6583 [3] (for carbon and low alloy steels).

Study of the Effects of Environment in the Fatigue Analysis on Existing LWR as Proposed in USNRC RG 1.207

2009 ◽  
Author(s):  
Eugene Tom ◽  
Milton Dong ◽  
Hong Ming Lee
Keyword(s):  
Fatigue Life ◽  
Correction Factor ◽  
Environmental Effects ◽  
Environmental Parameters ◽  
Computer Code ◽  
Cyclic Behavior ◽  
Simplified Approach ◽  
Asme Code ◽  
Detailed Computer ◽  
Fatigue Evaluation

US NRC Regulatory Guide 1.207 Rev. 0 provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light-water reactor (LWR) environment. Because of significant conservatism in quantifying other plant-related variables (such as cyclic behavior, including stress and loading rates) involved in cumulative fatigue life calculations, the design of the current fleet of reactors is satisfactory. For new plants under design and current operating plants considering applying for License Renewal, the environment effects may need to be considered in the design. RG 1.207 proposes using an environmental correction factor (Fen) to account for LWR environments by correcting the fatigue usage calculated with the ASME “air” curves. The Fen method is presented in NUREG/CR-6909, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials”. By definition, Fen is the ratio of fatigue life of the component material at room temperature air environments to its fatigue life in LWR coolant at operating temperature. To incorporate environmental effects into the fatigue evaluation, the fatigue usage is calculated using provisions set forth in Section III of the ASME Code, and is adjusted by multiplying a correction factor. The calculated Fen values are then used to incorporate environmental effects into ASME fatigue usage factor evaluation. Once the environmental correction factors have been determined, the previously calculated allowable number of cycles for each load set pair based on the current Code fatigue design curve can be adjusted to determine the new fatigue usage factors for environmental effects. This paper presents a study of the effect of the Regulatory Guide if it is to be implemented on the current fleet of LWR. A quick assessment of the sensitivity of the various environmental parameters is also included in this paper. The comparison of environmental effects between the simplified approach in this paper and the results with detailed computer analyses, such as Unisont’s propriety computer code UPIPENB (Ref. 4), will be our next research project to be presented in the future conference.

Prediction of Random High-Cycle Fatigue Life of LWR Components

1980 ◽  
Vol 102 (4) ◽  
pp. 378-386 ◽  
Author(s):  
Y. S. Shin
Keyword(s):  
Fatigue Life ◽  
Random Vibration ◽  
High Cycle Fatigue ◽  
State Of The Art ◽  
Random Excitation ◽  
Light Water Reactor ◽  
Cycle Fatigue ◽  
Flow Induced Vibration ◽  
Light Water ◽  
Water Reactor

This state-of-the-art review identifies and discusses the existing methods of predicting the high-cycle fatigue life, their limitations, and base-technology needs. The cycle stress-strain approach and the random vibration approach are reviewed, evaluated and discussed. It is applicable to estimating high-cycle fatigue damages of Light Water Reactor (LWR) components under the random excitation typical of flow-induced vibration.

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.

Evaluation of Fatigue Damage in LWR Water With and Without Threshold and Moderation Factor

2003 ◽  
Author(s):  
Makoto Higuchi ◽  
Kazuya Tsutsumi ◽  
Katsumi Sakaguchi
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Stainless Steels ◽  
Power Plants ◽  
Light Water Reactor ◽  
Low Alloy Steels ◽  
Light Water ◽  
Water Reactor ◽  
Fatigue Data ◽  
Fatigue Life Reduction

During the past twenty years, the fatigue initiation life of LWR structural materials, carbon, low alloy and stainless steels has been shown to decrease remarkably in the simulated LWR (light water reactor) coolant environments. Several models for evaluating the effects of environment on fatigue life reduction have been developed based on published environmental fatigue data. Initially, based on Japanese fatigue data, Higuchi and Iida proposed a model for evaluating such effects quantitatively for carbon and low alloy steels in 1991. Thereafter, Chopra et al. proposed other models for carbon, low alloy and stainless steels by adding American fatigue data in 1993. Mehta developed a new model which features the threshold concept and moderation factor in Chopra’s model in 1995. All these models have undergone various revisions. In Japan, the MITI (Ministry of International Trade and Industry) guideline on environmental fatigue life reduction for carbon, low alloy and stainless steels was issued in September 2000, for evaluating of aged light water reactor power plants. The MITI guideline provide equations for calculations applicable only to stainless steel in PWR water and consequently Higuchi et al. proposed in 2002 a revised model for stainless steel which incorporates new equations for evaluation of environmental fatigue reduction in BWR water. The paper compares the latest versions of these models and discusses the conservativeness of the models by comparison of the models with available test data.

scholarly journals Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

2017 ◽  
Vol 139 (6) ◽  
Author(s):  
O. K. Chopra ◽  
G. L. Stevens ◽  
R. Tregoning ◽  
A. S. Rao
Keyword(s):  
Fatigue Crack ◽  
Fatigue Life ◽  
Crack Initiation ◽  
Pressure Vessel ◽  
Fatigue Crack Initiation ◽  
Argonne National Laboratory ◽  
Light Water Reactor ◽  
Low Alloy Steels ◽  
Light Water ◽  
Water Reactor

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. However, the Code design curves do not address the effects of light water reactor (LWR) water environments. Existing fatigue strain-versus-life (ε–N) data illustrate significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory (Argonne) and elsewhere to investigate the effects of LWR environments on the fatigue life. This article summarizes the results of these studies. The existing fatigue ε–N data were evaluated to identify the various material, environmental, and loading conditions that influence the fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, Fen, which is the ratio of fatigue life in air at room temperature to the life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels (SSs), and nickel–chromium–iron (Ni–Cr–Fe) alloys and their weld metals. A review of the Code Section III fatigue adjustment factors of 2 and 20 is also presented, and the possible conservatism inherent in the choice is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation is presented.

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