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.

Revised Proposal of Fatigue Life Correction Factor Fen for Carbon and Low Alloy Steels in LWR Water Environments

2004 ◽  
Vol 126 (4) ◽  
pp. 438-444 ◽  
Author(s):  
Makoto Higuchi
Keyword(s):  
Fatigue Life ◽  
Correction Factor ◽  
Light Water Reactor ◽  
Low Alloy Steels ◽  
Light Water ◽  
High Temperature Water ◽  
Fatigue Data ◽  
Alloy Steels ◽  
Fatigue Life Reduction ◽  
Temperature Water

The fatigue life of carbon and low alloy steels decreases with reduction in strain rate in high temperature water such as in the case of a light water reactor coolant. The fatigue life reduction also depends on temperature and dissolved oxygen. The fatigue life correction factor Fen has been proposed as a method to assess the fatigue life reduction in such environments. Three different models for calculating Fen for carbon and low alloy steels have been proposed by Higuchi et al., Chopra et al., and Mehta. These models were compared using considerable environmental fatigue data that were tested and published in Japan and USA and piled up in the database “JNUFAD” by the author. These models give somewhat different results in the specific conditions and a revised model for calculating Fen is thus proposed by remedying the particular drawbacks of each. In this model, the same formula is used for carbon and low alloy steels and S*,T*,O*, and ε˙* are adopted in the formula after reevaluating every parameter. The revised proposal shows better correlation with the test data than the previous models.

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.

A Proposal of Fatigue Life Correction Factor Fen for Austenitic Stainless Steels in LWR Water Environments

2002 ◽  
Author(s):  
Makoto Higuchi ◽  
Kunihiro Iida ◽  
Akihiko Hirano ◽  
Kazuya Tsutsumi ◽  
Katsumi Sakaguchi
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Reduction Factor ◽  
New Models ◽  
Fatigue Data ◽  
Few Data ◽  
Fatigue Life Reduction ◽  
Remarkable Reduction

The fatigue life of austenitic stainless steel has recently been shown to undergo remarkable reduction with decrease in strain rate and increase in temperature in water. Either of these parameters as a factor of this reduction has been examined quantitatively and methods for predicting the fatigue life reduction factor Fen in any given set of conditions have been proposed. All these methods are based primarily on fatigue data in simulated PWR water owing to the few data available in simulated BWR water. Recent Japanese fatigue data in simulated BWR water clearly indicated the effects of the environment on fatigue degradation to be milder than under actual PWR conditions. A new method for determining Fen in BWR water was developed in the present study and a revised Fen in PWR water is also proposed based on new data. These new models differ from those previously used primarily with regard to the manner in which strain amplitude is considered to affect Fen in the environment.

A Proposal of Fatigue Life Correction Factor Fen for Austenitic Stainless Steels in LWR Water Environments

2003 ◽  
Vol 125 (4) ◽  
pp. 403-410 ◽  
Author(s):  
Makoto Higuchi ◽  
Kazuya Tsutsumi ◽  
Akihiko Hirano ◽  
Katsumi Sakaguchi
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Reduction Factor ◽  
New Models ◽  
Fatigue Data ◽  
Few Data ◽  
Fatigue Life Reduction ◽  
Remarkable Reduction

The fatigue life of austenitic stainless steel has recently been shown to undergo remarkable reduction with decrease in strain rate and increase in temperature in water. Either of these parameters as a factor of this reduction has been examined quantitatively and methods for predicting the fatigue life reduction factor Fen in any given set of conditions have been proposed. All these methods are based primarily on fatigue data in simulated PWR water owing to the few data available in simulated BWR water. Recent Japanese fatigue data in simulated BWR water clearly indicated the effects of the environment on fatigue degradation to be milder than under actual PWR conditions. A new method for determining Fen in BWR water was developed in the present study and a revised Fen in PWR water is also proposed based on new data. These new models differ from those previously used primarily with regard to the manner in which strain amplitude is considered to affect Fen in the environment.

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.

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.

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