Temperature Dependence of Reactor Water Environmental Fatigue Effects on Carbon, Low Alloy and Austenitic Stainless Steels

2008 ◽  
Author(s):  
William James O’Donnell ◽  
William John O’Donnell
Keyword(s):  
Environmental Effects ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Design Criteria ◽  
Strain Rates ◽  
Worst Case ◽  
Reactor Water ◽  
Fatigue Design ◽  
Fatigue Data ◽  
Water Effects

Recent studies of the environmental fatigue data for carbon, low alloy and austenitic stainless steels have shown that reactor water effects are significantly less deleterious as temperatures are reduced below 350 °C (662 °F). At temperatures below 150 °C (302 °F) the reduction in life due to reactor water environmental effects is less than a factor of 2, and the existing ASME Code Section III fatigue design curves for air can be used. The latter include a factor of 20 on cycles whereas the ASME Subgroup on Fatigue Strength (SGFS) has determined that a factor of 10 should be used on the mean failure curves which include reactor water effects. These factors account for scatter in the data, surface finish effects, size effects, and environmental effects. Reactor water environmental degradation dependence on temperature is determined using variations of the statistical models developed by Chopra and Shack, Higuchi, Iiada, Asada, Nakamura, Van Der Sluys, Yukawa, Mehta, Leax and Gosselin, References [1 through 22]. Comparisons of the resulting proposed environmental fatigue design criteria with reactor water environmental fatigue data are made. These comparisons show that the Code factors of 2 and 20 on stress and cycles are maintained for air environments, and the 2 and 10 Code factors are maintained for the reactor water environments. Environmental fatigue criteria are given for both worst case strain rates and for arbitrary strain rates. These design criteria do not require the designer to consider sequence of loading, hold times, transient rates, and other operating details which may change during 60 years of plant operation.

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.

Derivation of Design Fatigue Curves for Austenitic Stainless Steel Grades 1.4541 and 1.4550 Within the German Nuclear Safety Standard KTA 3201.2

2013 ◽  
Author(s):  
Xaver Schuler ◽  
Karl-Heinz Herter ◽  
Jürgen Rudolph
Keyword(s):  
Austenitic Stainless Steel ◽  
Stainless Steels ◽  
Room Temperature ◽  
Elevated Temperatures ◽  
Austenitic Stainless Steels ◽  
Nuclear Safety ◽  
Safety Standard ◽  
Fatigue Design ◽  
Steel Grades ◽  
Best Fit

Titanium and niobium stabilized austenitic stainless steels X6CrNiTi18-10S (material number 1.4541, correspondent to Alloy 321) respectively X6CrNiNb18-10S (material number 1.4550, correspondent to Alloy 347) are widely applied materials in German nuclear power plant components. Related requirements are defined in Nuclear Safety Standard KTA 3201.1. Fatigue design analysis is based on Nuclear Safety Standard KTA 3201.2. The fatigue design curve for austenitic stainless steels in the current valid edition of KTA 3201.2 is essentially identical with the design curve included in ASME-BPVC III, App I (ed. 2007, add. July 2008 respectively back editions). In the current code revision activities of KTA 3201.2 the compatibility of latest in air fatigue data for austenitic stainless steels with the above mentioned grades were examined in detail. The examinations were based on statistical evaluations of 149 strain controlled test data at room temperature and 129 data at elevated temperatures to derive best-fit mean data curves. Results of two additional load controlled test series (at room temperature and 288°C) in the high cycle regime were used to determine a technical endurance limit at 107 cycles. The related strain amplitudes were determined by consideration of the cyclic stress strain curve. The available fatigue data for the two austenitic materials at room temperature and elevated temperatures showed a clear temperature dependence in the high cycle regime demanding for two different best-fit curves. The correlation of the technical endurance limit(s) at room temperature and elevated temperatures with the ultimate strength of the materials is discussed. Design fatigue curves were derived by application of the well known factors to the best-fit curves. A factor of SN = 12 was applied to load cycles correspondent to the NUREG/CR-6909 approach covering influences of data scatter, surface roughness, size and sequence. In terms of strain respectively stress amplitudes in the high cycle regime, for elevated temperatures (>80°C) a factor of Sσ = 1.79 was applied considering and combining in detail the partial influences of data scatter surface roughness, size and mean stress. For room temperature a factor of Sσ = 1.88 shall be applied. As a result, new design fatigue curves for austenitic stainless steel grades 1.4541 and 1.4550 will be available within the German Nuclear Safety Standard KTA 3201.2. The fatigue design rules for all other austenitic stainless steel grades will be based on the new ASME-BPVC III, App I (ed. 2010) design curve.

Development of New Design Fatigue Curves in Japan: Proposal of a New Fatigue Evaluation Method

2018 ◽  
Author(s):  
Seiji Asada ◽  
Akihiko Hirano ◽  
Toshiyuki Saito ◽  
Yasukazu Takada ◽  
Hideo Kobayashi
Keyword(s):  
Stainless Steels ◽  
Evaluation Method ◽  
Austenitic Stainless Steels ◽  
Fatigue Tests ◽  
Carbon Steels ◽  
Low Alloy Steels ◽  
Stress Effect ◽  
Fatigue Data ◽  
Alloy Steels ◽  
Fatigue Evaluation

In order to develop new design fatigue curves for carbon steels & low-alloy steels and austenitic stainless steels and a new design fatigue evaluation method that are rational and have clear design basis, Design Fatigue Curve (DFC) Phase 1 subcommittee and Phase 2 subcommittee were established in the Atomic Energy Research Committee in the Japan Welding Engineering Society (JWES). The study on design fatigue curves was actively performed in the subcommittees. In the subcommittees, domestic and foreign fatigue data of small test specimens in air were collected and a comprehensive fatigue database (≈6000 data) was constructed and the accurate best-fit curves of carbon steels & low-alloy steels and austenitic stainless steels were developed. Design factors were investigated. Also, a Japanese utility collaborative project performed large scale fatigue tests using austenitic stainless steel piping and low-alloy steel flat plates as well as fatigue tests using small specimens to obtain not only basic data but also fatigue data of mean stress effect, surface finish effect and size effect. Those test results were provided to the subcommittee and utilized the above studies. Based on the above studies, a new fatigue evaluation method has been developed.

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.

Defect Microstructures and Deformation Mechanisms in Irradiated Austenitic Stainless Steels

1996 ◽  
Vol 439 ◽  
Author(s):  
S. M. Bruemmer ◽  
J. I. Cole ◽  
R. D. Carter ◽  
G. S. Was
Keyword(s):  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Heavy Ion ◽  
Deformation Mechanisms ◽  
Strain Rates ◽  
Dislocation Loops ◽  
Intergranular Cracking ◽  
Localized Deformation ◽  
Electron Microscopies ◽  
Defect Microstructures

AbstractMicrostructural evolution and deformation behavior of austenitic stainless steels are evaluated for neutron, heavy-ion and proton irradiated materials. Radiation hardening in austenitic stainless steels is shown to result from the evolution of small interstitial dislocation loops during lightwater-reactor (LWR) irradiation. Available data on stainless steels irradiated under LWR conditions have been analyzed and microstructural characteristics assessed for the critical fluence range (0.5 to 10 dpa) where irradiation-assisted stress corrosion cracking susceptibility is observed. Heavy-ion and proton irradiations are used to produce similar defect microstructures enabling the investigation of hardening and deformation mechanisms. Scanning electron, atomic force and transmission electron microscopies are employed to examine tensile test strain rate and temperature effects on deformation characteristics. Dislocation loop microstructures are found to promote inhomogeneous planar deformation within the matrix and regularly spaced steps at the surface during plastic deformation. Twinning is the dominant deformation mechanism at rapid strain rates and at low temperatures, while dislocation channeling is favored at slower strain rates and at higher temperatures. Both mechanisms produce highly localized deformation and large surface slip steps. Channeling, in particular, is capable of creating extensive dislocation pileups and high stresses at internal grain boundaries which may promote intergranular cracking.

EDF/NRC High-Cycle Fatigue Database Proposal

2015 ◽  
Author(s):  
T. P. Métais ◽  
G. Stevens ◽  
G. Blatman ◽  
J. C. Le Roux ◽  
R. L. Tregoning
Keyword(s):  
Stainless Steels ◽  
Data Exchange ◽  
High Cycle Fatigue ◽  
Fatigue Testing ◽  
Austenitic Stainless Steels ◽  
Nuclear Regulatory Commission ◽  
Fatigue Curve ◽  
Research Data ◽  
Cycle Fatigue ◽  
Fatigue Data

Revised fatigue curves for austenitic stainless steels are currently being considered by several organizations in various countries, including Japan, South Korea, and France. The data available from laboratory tests indicate that the mean air curve considering all available austenitic material fatigue data may be overly conservative compared to a mean curve constructed from only those data representative of a particular type of material. In other words, developing separate fatigue curves for each of the different types of austenitic materials may prove useful in terms of removing excess conservatism in the estimation of fatigue lives. In practice, the fatigue curves of interest are documented in the various international design codes. For example, in the 2009 Addenda of Section III of the ASME Boiler and Pressure Vessel (BPV) Code [1], a revised design air fatigue curve for austenitic materials was implemented that was based on NRC research models [2]. More recently, in Japan, various industrial groups have joined their efforts to create the Design Fatigue Curve Sub-committee (DCFS) with the objective to reassess the fatigue curves [3]. In France, EDF/AREVA and CEA are developing a new fatigue curve for austenitic stainless steels [4]. More specifically, in 2014, EDF presented a paper on high-cycle fatigue analysis which demonstrated that the factor on the strain amplitude could be reduced from 2 to 1.4 for the RCC-M austenitic stainless steel grades [5]. Recently, discussions between EDF and the U.S. Nuclear Regulatory Commission (NRC) have led both parties to recognize that there is a need to exchange worldwide research data from fatigue testing to promote a common, vetted database available to all researchers. These discussions have led EDF and NRC to pursue a collaborative agreement and associated fatigue data exchange, with the intent to assemble all available fatigue data for austenitic materials into a standardized format. The longer term objective is to perform common analyses on the consolidated set of data. This paper summarizes the intent and of the preliminary results of this cooperation and also provides insights from both organizations on possible future activities and participation in the global exchange of fatigue research data.

scholarly journals Development of a Water Environment Fatigue Design Curve for Austenitic Stainless Steels

2003 ◽  
Author(s):  
Thomas R. Leax
Keyword(s):  
Stainless Steels ◽  
Fatigue Behavior ◽  
Water Environment ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Light Water Reactor ◽  
Design Curve ◽  
Fatigue Design ◽  
American Society ◽  
Design Factors

Technical support is provided for a fatigue curve that could potentially be incorporated into Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This fatigue curve conservatively accounts for the effects of light water reactor environments on the fatigue behavior of austenitic stainless steels. This paper presents the data, statistical methods, and basis for the design factors appropriate for Code applications. A discussion of the assumptions and methods used in design curve development is presented.

Sign in / Sign up

Export Citation Format

Share Document