Pressurized Water Reactor Environment Effect on 316 Stainless Steel Stress Hardening/Softening: An Experimental Study

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Stress Strain ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Light Water ◽  
Water Reactor ◽  
Life Estimation ◽  
316 Stainless Steel ◽  
Water Reactors

In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S∼N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

Low Cycle Fatigue Life of Type 316 Stainless Steel Notched Specimen Under Proportional and Non-Proportional Multiaxial Loadings

2010 ◽  
Author(s):  
Takamoto Itoh
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Strain Path ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue ◽  
Notched Specimen ◽  
Life Estimation ◽  
316 Stainless Steel ◽  
Strain Parameter ◽  
Strain Paths

This study discusses multiaxial low cycle fatigue life of notched specimen under proportional and non-proportional loadings at room temperature. Strain controlled multiaxial low cycle fatigue tests were carried out using smooth and circumferentially notched round-bar specimens of type 316 stainless steel. Four kinds of notched specimens were employed of which elastic stress concentration factors, Kt, are 1.5, 2.5, 4.2 and 6.0. The strain paths include proportional and non-proportional loadings. The former employed a push-pull straining or a reversed torsion straining. The latter was achieved by strain path where axial and shear strains has 90 degree phase difference but their amplitudes is the same based on von Mises’ criterion. The notch dependency of multiaxial low cycle fatigue life and the life estimation are discussed. The lives depend on both Kt and strain path. The strain parameter for the life estimation is also discussed with the non-proportional strain parameter proposed by the author with introducing Kt. The proposed parameter gives a satisfactory correlation with multiaxial low cycle fatigue life of notched specimen of type 316 stainless steel under proportional and non-proportional loadings.

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.

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.

Mean Stress Effect on Fatigue Properties of Type 316 Stainless Steel in Pressurized Water Reactors Primary Water Environment

2019 ◽  
Vol 141 (5) ◽  
Author(s):  
Masayuki Kamaya
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Water Environment ◽  
Stress Effect ◽  
Mean Stress ◽  
Pressurized Water ◽  
316 Stainless Steel ◽  
Mean Stress Effect ◽  
Primary Water ◽  
The Mean

The mean stress effect on the fatigue life of type 316 stainless steel was investigated in simulated pressurized water reactor (PWR) primary water and air at 325 °C. The tests in air environment have revealed that the fatigue life was increased with application of the positive mean stress for the same stress amplitude because the strain range was decreased by hardening of material caused by increased maximum peak stress. On the other hand, it has been shown that the fatigue life obtained in simulated PWR primary water was decreased compared with that obtained in air environment even without the mean stress. In this study, type 316 stainless steel specimens were subjected to the fatigue test with and without application of the positive mean stress in high-temperature air and PWR water environments. First, the mean stress effect was discussed for high-temperature air environment. Then, the change in fatigue life in the PWR water environment was evaluated. It was revealed that the change in the fatigue life due to application of the mean stress in the PWR water environment could be explained in the same way as for the air environment. No additional factor was induced by applying the mean stress in the PWR water environment.

scholarly journals Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

2021 ◽  
Vol 2 (4) ◽  
pp. 318-335
Author(s):  
Kang-Seog Kim ◽  
William A. Wieselquist
Keyword(s):  
Nuclear Data ◽  
Data File ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
Benchmark Assessments ◽  
Light Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Data Library ◽  
Neutronic Characteristics

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

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