Fatigue Strain–Life Behavior of Austenitic Stainless Steels in Pressurized Water Reactor Environments

2015 ◽  
Author(s):  
Paul Wilhelm ◽  
Paul Steinmann ◽  
Jürgen Rudolph
Keyword(s):  
Fatigue Life ◽  
Stainless Steels ◽  
National Laboratory ◽  
Reference Conditions ◽  
Argonne National Laboratory ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

A statistical model for austenitic stainless steels for predicting the effect of pressurized water reactor environments on fatigue life for a range of temperatures and strain rates is developed based on analysis of available material data. The compiled fatigue curve data include not only results from America (Keller (1971), Conway (1975), Hale (1977), and Argonne National Laboratory (1999–2005)), but also from Europe (Solin (2006), Le Duff (2008–2010), De Baglion (2011, 2012), Huin (2013) …) and Japan (Kanasaki (1997)). Only fatigue data from polished specimens of wrought material tested under strain control were considered; hollow specimens were not treated herein. The fatigue life correction factor used in this paper was defined as the ratio of life in water at 300 °C (reference conditions) to that in water at service conditions. The model is recommended for predicting fatigue lives that are 103–105 cycles.

Effects of Pressurized Water Reactor Medium on the Fatigue Life of Austenitic Stainless Steels

2015 ◽  
Vol 137 (6) ◽  
Author(s):  
Paul Wilhelm ◽  
Jürgen Rudolph ◽  
Paul Steinmann
Keyword(s):  
Fatigue Life ◽  
Stainless Steels ◽  
Reference Conditions ◽  
Austenitic Stainless Steels ◽  
Strain Rates ◽  
Correction Factors ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Service Conditions

A statistical model for austenitic stainless steels for predicting the effect of pressurized water reactor (PWR) environments on fatigue life for a range of temperatures and strain rates is developed based on analysis of available material data from USA, Europe, and Japan. Only fatigue data from polished specimens of wrought material tested under strain control were considered. Hollow specimens were not treated in the final calculations. The fatigue life correction factors were defined as the ratio of life in water at 300 °C (572 °F) (reference conditions) to that in water at service conditions. The model is recommended for predicting fatigue lives that are 103–105 cycles.

Crack Growth Behavior of Nickel Alloy Welds in a PWR Environment

2007 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Crack Growth ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Crack Growth Behavior ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Ni Alloy ◽  
The Individual

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

Discussion of Fatigue Data for Austenitic Stainless Steels

2014 ◽  
Author(s):  
Paul Wilhelm ◽  
Paul Steinmann ◽  
Jürgen Rudolph
Keyword(s):  
Fatigue Life ◽  
Stainless Steels ◽  
National Laboratory ◽  
Hold Time ◽  
Argonne National Laboratory ◽  
Austenitic Stainless Steels ◽  
Reduction Factor ◽  
Fatigue Model ◽  
Fatigue Data ◽  
Number Of Cycles

The first results of a detailed fatigue model for austenitic stainless steels in general and for the grades 1.4541 and 1.4550 are presented to describe the effect of the light water reactor (LWR) coolant environments on the fatigue life. The statistical evaluations are based on strain (and load) controlled test series from different institutions. The compiled fatigue data include not only results from America (Keller (1971), Conway (1975), Hale (1977), and Argonne National Laboratory (ANL)(1999–2005)), but also from Europe (Solin (2006), Le Duff (2008–2010), De Baglion (2011, 2012), Huin (2013),…) and Japan (Kanasaki (1997)). The fatigue life is defined as the number of cycles necessary for tensile stress to drop 25 percent from its peak value. Fatigue lives defined by other failure criteria are normalized to the load reduction of 25 percent, before the statistical analysis is performed. The fatigue data are expressed in terms of the Langer equation and the parameter “material variability and data scatter” is quantified. Additionally, fatigue data in air of roughened specimens are compiled and discussed. A reduction factor of 2.5 on number of cycles is derived to cover the maximum allowed surface roughness. Based on the derived best-fit curves, design-curves in air and, in a second step, environmentally assisted fatigue (EAF) curves for LWR environments, which consider temperature, strain rate, dissolved oxygen content, and hold-time effects, will be incorporated in the detailed fatigue model in the future.

Effects of Boiling Water Reactor Medium on the Fatigue Life of Austenitic Stainless Steels

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Paul Wilhelm ◽  
Paul Steinmann ◽  
Jürgen Rudolph
Keyword(s):  
Fatigue Life ◽  
Stainless Steels ◽  
Reference Conditions ◽  
Austenitic Stainless Steels ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Strain Rates ◽  
Correction Factors ◽  
Water Reactor ◽  
Service Conditions

A statistical model for austenitic stainless steels (SSs) for predicting the effect of boiling water reactor (BWR) environments on fatigue life for a range of temperatures and strain rates is developed based on the analysis of available material data from Europe, U.S., and Japan. Only fatigue data from polished specimens of wrought material tested under strain control were considered. Hollow specimens were not treated in the final calculations. The fatigue-life correction factors were defined as the ratio of life in water at 240 °C (464 °F) (reference conditions) to that in water at service conditions. The model is recommended for predicting fatigue lives that are 102–104 cycles.

Technical Basis for Revision of Code Case N-809 on Reference Fatigue Crack Growth Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments

2021 ◽  
Author(s):  
Russell C. Cipolla ◽  
Warren H. Bamford ◽  
Kiminobu Hojo ◽  
Yuichiro Nomura
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Stainless Steels ◽  
Growth Curves ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
R Ratio ◽  
Technical Basis

Abstract Reference fatigue crack growth curves for austenitic stainless steels exposed to pressurized water reactor environments have been available in the ASME Code, Section XI in their present form with the publication of Code Case N-809 in Supplement 2 to the 2015 Code Edition. The reference curves are dependent on temperature, loading rate (loading rise time), mean stress (R-ratio), and cyclic stress intensity factor range (ΔK), which are all contained in the model. Since the first implementation of this Code Case, additional data have become available, and the purpose of this paper is to provide the technical basis for revision of the Code Case. Changes have been made in three areas: R-ratio behavior, threshold for crack growth (ΔKth), and crack growth rate dependence on ΔK. In addition, the temperature model was revisited to study the temperature effects for T < 150°C, where the current model predicts an increase in da/dN based on limited test data at about 100°C (200°F). At this point, the current temperature model is considered conservative and no change is proposed in this revision to N-809. The R-ratio model has been revised for both high and low carbon stainless steels, a significant improvement over the original procedures. Perhaps the most important revision is in the area of the threshold for the initiation of fatigue crack growth; such data are difficult to obtain, and the previous model was very conservative. Finally, the crack growth exponent was revised slightly to make it consistent with the regression analysis of the original data.

Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China

2010 ◽  
Author(s):  
Yan Wang ◽  
Xie Heng
Keyword(s):  
Uncertainty Analysis ◽  
System Analysis ◽  
High Efficiency ◽  
Current System ◽  
National Laboratory ◽  
Pressurized Water Reactor ◽  
Economic Competition ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

scholarly journals Fatigue Strength Evaluation Of Pressurizer Wall Structure In Pressurized Water Reactor

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Roziq Himawan
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
System Analysis ◽  
Fatigue Curve ◽  
Failure Criteria ◽  
Wall Structure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Design Qualification

<p>Fatigue strength evaluations have been performed to the pressurizer component in Pressurized Water Reactor. Fatigue is the main failure mechanism of material during system in operation. Therefore, this evaluation becomes important to be performed since the pressurizer has a very important function in the reactor’s system. Analysis was performed by using Nuclear Power Plant operation data from 40 years operation and base on Miner theory. This analysis covered all stress level experienced by the reactor during the service. To determine the value of fatigue usage factor a, fatigue curve of SA 533 material was applied. Analysis results show that the cumulative fatigue damage during 40 years in operation is 4,23×10<sup>-4</sup>. This value still far enough below failure criteria, which a value is 1. Therefore, the pressurizer design has already fulfilled the design qualification in term of fatigue aspect.</p>

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