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.

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.

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.

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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