A Review of the Effects of Coolant Environments on the Fatigue Life of LWR Structural Materials

2009 ◽  
Vol 131 (2) ◽  
Author(s):  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Fatigue Life ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Crack Initiation ◽  
Austenitic Stainless Steels ◽  
Structural Materials ◽  
Low Alloy Steels ◽  
Asme Code ◽  
Alloy Steels

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies design curves for the fatigue life of structural materials in nuclear power plants. However, the effects of light water reactor (LWR) coolant environments were not explicitly considered in the development of the design curves. The existing fatigue-strain-versus-life (ε-N) data indicate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. Under certain environmental and loading conditions, fatigue lives in water relative to those in air can be a factor of 15 lower for austenitic stainless steels and a factor of ≈30 lower for carbon and low-alloy steels. This paper reviews the current technical basis for the understanding of the fatigue of piping and pressure vessel steels in LWR environments. The existing fatigue ε-N data have been evaluated to identify the various material, environmental, and loading parameters that influence fatigue crack initiation and to establish the effects of key parameters on the fatigue life of these steels. Statistical models are presented for estimating fatigue life as a function of material, loading, and environmental conditions. An environmental fatigue correction factor for incorporating the effects of LWR environments into ASME Code fatigue evaluations is described. This paper also presents a critical review of the ASME Code fatigue design margins of 2 on stress (or strain) and 20 on life and assesses the possible conservatism in the current choice of design margins.

Margins for ASME Code Fatigue Design Curve: Effects of Surface Finish and Material Variability

2003 ◽  
Author(s):  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Environmental Parameters ◽  
Light Water Reactor ◽  
Code Design ◽  
Low Alloy Steels ◽  
Design Curve ◽  
Fatigue Design ◽  
Asme Code ◽  
Alloy Steels

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. This Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life (ε-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue ε-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue ε-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for the ASME Code fatigue design curve is presented.

Development of a Fatigue Design Curve for Austenitic Stainless Steels in LWR Environments: A Review

2002 ◽  
Author(s):  
Omesh K. Chopra
Keyword(s):  
Stainless Steels ◽  
Pressure Vessel ◽  
Type Strain ◽  
Nuclear Power ◽  
Austenitic Stainless Steels ◽  
Environmental Parameters ◽  
Light Water Reactor ◽  
Code Design ◽  
Fatigue Design ◽  
Asme Code

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components and specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, and flow rate, on the fatigue lives of these steels are summarized. Statistical models are presented for estimating the fatigue ε–N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue design curves.

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.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Evaluation of P-T Limit Curves According to Alternative Fracture Toughness Requirements

2014 ◽  
Vol 1051 ◽  
pp. 896-901
Author(s):  
Sin Ae Lee ◽  
Sung Jun Lee ◽  
Sang Hwan Lee ◽  
Yoon Suk Chang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Ductile Failure ◽  
Evaluation Procedure ◽  
Limit Curve ◽  
Asme Code ◽  
Temperature And Pressure ◽  
Reactor Pressure

During the heat-up and cool-down processes of nuclear power plants, temperature and pressure histories are to be maintained below the P-T limit curve to prevent the non-ductile failure of the RPV(Reactor Pressure Vessel). The ASME Code Sec. XI, App. G describe the detailed procedure for generating the P-T limit curve. The evaluation procedure is containing the evaluation methods of RTNDT using 10CFR50.61. However, recently, Alternative fracture toughness requirements were released 10CFR50.61a. Therefore, in this study, RTNDT of RPV according to the 10CFR50.61a was calculated and used for evaluation of P-T limit curve of a typical RPV under cool-down condition. As a result, it was proven that the P-T curve obtained from 10CFR50.61 is conservative because RTNDT value obtained from the alternative fracture toughness requirements are significantly low.

Observation of Fatigue Crack Initiation and Propagation for Carbon and Low-Alloy Steels in Oxygenated Water at Elevated Temperature

2006 ◽  
Author(s):  
Makoto Higuchi ◽  
Katsumi Sakaguchi
Keyword(s):  
Fatigue Crack ◽  
Fatigue Life ◽  
Elevated Temperature ◽  
Crack Initiation ◽  
Fatigue Crack Initiation ◽  
Low Alloy Steels ◽  
Test Conditions ◽  
Alloy Steels ◽  
Crack Initiation Life ◽  
Temperature Water

Low cycle fatigue life of structural materials in LWR plants decreases remarkably in elevated temperature water depending on strain rate, temperature, water chemistry and material properties. The maximum reduction rate in fatigue life for carbon and low alloy steels is over 100 in severe conditions. Fatigue life is composed of fatigue crack initiation life and consequent propagation life. It is important to know the proportion of crack initiation life to propagation life in water environment when developing a model to estimate fatigue crack initiation life. The beachmark imprinting method was used to monitor fatigue crack initiation and consequent propagation. Environmental test conditions varied widely from severely accelerated conditions of high temperature and dissolved oxygen to mild conditions of lower temperature and oxygen. Fatigue crack initiation life could be determined using the beachmark imprinting method for all test conditions. Based on obtained test results, the susceptibility of each parameter in NWC and the relationships between NWC/NW and environmental fatigue life correction factor Fen under various conditions are discussed, but a good relationship could not be detected due to widely scattered data and a model to predict fatigue crack initiation life could not be proposed.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

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.

Hydrazine-Dependency of Low-Alloy Steel Flow-Accelerated Corrosion in a Deoxygenated Solution at 250°C

2007 ◽  
Vol 26-28 ◽  
pp. 1133-1136
Author(s):  
Ki Woung Sung ◽  
Hyun Il Seo ◽  
Uh Chul Kim ◽  
Wan Young Maeng
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Alloy Steels ◽  
Accelerated Corrosion ◽  
Test Loop ◽  
Concentration Dependency ◽  
Alloy Steels ◽  
Steel Flow ◽  
Flow Accelerated Corrosion ◽  
Experimental Weight Loss

In the nuclear power plants (NPPs), wall thinning of the piping materials is generally caused by a flow-accelerated corrosion (FAC) and leads to a rupture with no warning unless it is detected and repaired in a timely manner. To reduce the FAC, it is better to use low-alloy steels, such as 1Cr-½Mo and 2¼Cr-1Mo, having higher FAC-resistance than carbon steel. Meanwhile, in the secondary water chemistry at the NPPs in Korea, hydrazine concentration is maintained within the range of 100~150 ppb. For applying these low-alloys to a piping material, we investigated the influence of hydrazine concentration on their FAC. An experiment was carried out at pH25°C of 9 controlled with ammonia in a deoxygenated aqueous solution containing 0~250 ppb-hydrazine by using a FAC test loop at 250°C for 300 hours. Experimental weight loss showed a hydrazine concentration dependency of the FAC in this concentration range, giving minimum at 150 ppb.

Modifications of the 2016 Edition of the RCC-M Code to Account for Environmentally Assisted Fatigue

2016 ◽  
Author(s):  
Stéphan Courtin ◽  
Thomas Métais ◽  
Manuela Triay ◽  
Eric Meister ◽  
Stéphane Marie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Steering Committee ◽  
Nuclear Industry ◽  
Life Extension ◽  
Asme Code ◽  
Nickel Base ◽  
Fatigue Evaluation ◽  
International Experts

The French nuclear industry has to face nowadays a series of challenges it did not have to face a decade ago. The most significant one is to ensure a reliable and safe operation of Nuclear Power Plants (NPP) in a context of both an ageing reactor fleet and new builds. The new constructions need rules that integrate a strong operation feedback while the older NPPs need rules that will guarantee the life extension beyond 40 years of operation. In this context, a new edition of the French RCC-M Code is planned for 2016. This new edition integrates the modifications made to the Code as a result of Requests for Modification (RM), which can be submitted by anyone and which help to continuously improve the quality and robustness of the Code. Concerning fatigue analyses, the RCC-M Code steering committee has acknowledged end of 2014 the reception of two RM to modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as to integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. The contents of these two RM were based on the proposals presented in Reference [1]. AFCEN required a technical review of these two RM and this task was performed by a working group composed by French and international experts. This process concluded to the approval of these two RM to be integrated to the 2016 edition of the RCC-M Code. This paper offers a presentation of these two new Rules in Probation Phase (RPP), this format being quite similar to Code Cases proposed by ASME Code.

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