Material and Temperature Effects in Low and High Cycle EAF of Austenitic Stainless Steels

2021 ◽  
Author(s):  
Tommi Seppänen ◽  
Jouni Alhainen ◽  
Esko Arilahti ◽  
Jussi Solin
Keyword(s):  
Nuclear Power ◽  
Austenitic Stainless Steels ◽  
Reference Curve ◽  
Fatigue Design ◽  
Equivalent Number ◽  
Specific Strain ◽  
Simple Change ◽  
The Mean ◽  
Number Of Cycles ◽  
Best Fit

Abstract Fatigue design of nuclear power plant pressure boundary components necessitates the use of design curves, where the allowable number of cycles is a function of the applied stress intensity. Design curves are derived from best-fit curves to small-specimen data, which comprises a range of materials, heats, temperatures and test techniques. This paper continues the series of papers most recently published in PVP2020-21136. At VTT, heat specific strain-life data has been obtained. In this paper, using our data and literature data we demonstrate the significance these heat and temperature specific best-fit curves can have on the predicted number of cycles, when also considering the detrimental effect of environment through Fen factors. Example calculations show that in most cases, a simple change of the reference curve from the Code best-fit curve to a more realistic one adds considerably to the number of predicted cycles, or alternatively, reduces the cumulative usage at equivalent number of cycles. Mainly for high cycle fatigue at high temperature, best-fit heat-specific curves may lie below the reference, taken as the mean curve in NUREG/CR-6909. However, refinement of calculation criteria and/or the Fen methodology is considered to provide additional relief to these cases so that the cumulative usage factor calculation can still be kept below unity.

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.

Statistical Analyses of High Cycle Fatigue French Data for Austenitic SS

2014 ◽  
Author(s):  
Géraud Blatman ◽  
Thomas Métais ◽  
Jean-Christophe Le Roux ◽  
Simon Cambier
Keyword(s):  
Strain Amplitude ◽  
High Cycle Fatigue ◽  
Experimental Testing ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Statistical Analyses ◽  
Starting Point ◽  
The Mean ◽  
Number Of Cycles ◽  
Selection Of

In the 2009 version of the ASME BPV Code, a set of new design fatigue curves were proposed to cover the various steels of the code. These changes occurred in the wake of publications [1] showing that the mean air curve used to build the former ASME fatigue curve did not always represent accurately laboratory results. The starting point for the methodology to build the design curve is the mean air curve obtained through laboratory testing: coefficients are then applied to the mean air curve in order to bridge the gap between experimental testing and reactor conditions. These coefficients on the number of cycles and on the strain amplitude are equal to 12 and 2 respectively in the 2009 ASME BPV code, using the mean air curve proposal from NUREG/CR-6909 [1]. Internationally, with the same mean air curve, other proposals have emerged and especially in France [2]-[3] where a consensus seems to be reached on the reduction of the coefficient on strain amplitude. This paper provides statistical analyses of the experimental data obtained in France at high-cycle for austenitic stainless steels. It enables to bring arguments for the selection of a coefficient on strain amplitude in the French RCC-M code, where less scatter on the data is witnessed due to fewer material grades.

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.

Evaluation of Low-Cycle Fatigue in Simulated PWR Environment

2006 ◽  
Vol 326-328 ◽  
pp. 1011-1014 ◽  
Author(s):  
Ill Seok Jeong ◽  
Sang Jai Kim ◽  
Taek Ho Song ◽  
Sung Yull Hong
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Water Environment ◽  
Fatigue Tests ◽  
Cycle Fatigue ◽  
Fatigue Design ◽  
Number Of Cycles ◽  
Measured Length

For developing fatigue design curve of cast stainless steel that is used in piping material of nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical solid fatigue specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with other previous results.

Development of Fatigue Design Curves for Pressure Vessel Alloys Using a Modified Langer Equation

1979 ◽  
Vol 101 (4) ◽  
pp. 292-297 ◽  
Author(s):  
D. R. Diercks
Keyword(s):  
Pressure Vessel ◽  
Room Temperature ◽  
Low Cycle Fatigue ◽  
Austenitic Stainless Steels ◽  
Cyclic Stress ◽  
Allowable Stress ◽  
Fitting Procedure ◽  
Fatigue Design ◽  
Alloy 800 ◽  
Best Fit

The Jaske and O’Donnell [1] curve-fitting procedure for analyzing fatigue data generated between room temperature and 427° C (800° F) for several pressure vessel alloys is reexamined in the present paper. Substantial improvements over their best-fit curves to the data are found to result from two proposed modifications to their procedure, namely 1) the use of a variable exponent in the Langer equation, and 2) minimization of the sum of the squares of the errors in the logarithms of the cyclic-stress amplitudes rather than in the stress amplitudes directly. Likewise, important differences are observed for the resultant allowable stress-amplitude values for design purposes. In particular, the present analysis permits higher allowable stress amplitudes in the critical low-cycle fatigue-life region for the austenitic stainless steels, alloy 800, and alloy 600. Two best-fit curves and the associated sets of allowable stress amplitudes, corresponding to the inclusion or deletion of load-controlled data, are obtained for alloy 718.

Proposed ASME Code High Cycle Fatigue Design Curves for Austenitic and Ferritic Steels

2017 ◽  
Author(s):  
Sampath Ranganath ◽  
Hardayal S. Mehta ◽  
Nathan A. Palm ◽  
John Hosler
Keyword(s):  
High Cycle Fatigue ◽  
Ferritic Steels ◽  
Mean Stress ◽  
Cycle Fatigue ◽  
Design Curve ◽  
Fatigue Design ◽  
Asme Code ◽  
The Mean ◽  
Number Of Cycles ◽  
Fatigue Evaluation

The ASME Code fatigue curves (S–N curves) are used in the fatigue evaluation of reactor components. For the assessment of high frequency cyclic loading (such as those produced by flow-induced vibrations), where the number of cycles is expected to be very large and cannot be estimated, the stresses are evaluated by comparison with the fatigue limit1 at 1011 cycles. Other high cycle events of finite time duration (e.g. safety relief loading), where the number of cycles is large but well defined, the fatigue evaluation is performed by comparing the calculated stress with the allowable values defined by the high cycle fatigue design curve. This paper discusses the development of fatigue design curves for austenitic and ferritic steels when the number of cycles is in the range 106 – 1011 cycles. The first part of the paper addresses austenitic stainless steel components which are used for reactor internals. Specifically, the approach described here uses temperature dependent properties (cyclic yield strength, cyclic ultimate strength) for the mean stress correction and the correction for the modulus of elasticity. The high cycle fatigue design curve is developed by applying the mean stress and the E correction on the reversing load mean data curve and applying a factor of 2 on stress. The generic methodology developed for austenitic steel was applied to carbon and low alloy steels also. The proposed fatigue design curves are part of a draft ASME Code Case being considered by the ASME Code Subgroup on Design Methods. This paper describes the technical basis for the proposed ASME Code Case for the high cycle fatigue design curves for austenitic and ferritic steels.

Fatigue Benchmark Comparison Effort Between Code_Aster and CNNC/NPIC Software: Part 2

2020 ◽  
Author(s):  
Hai Xie ◽  
Zichen Kong ◽  
Xuejiao Shao ◽  
Tanguy Mathieu ◽  
Furui Xiong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Combination Method ◽  
Practical Implementation ◽  
Phase 2 ◽  
Software Comparison ◽  
Fatigue Design ◽  
Nickel Base ◽  
Fatigue Evaluation

Abstract Fatigue is identified as a significant degradation mode that affects nuclear power plants world-wide. Recent research on the interaction between fatigue degradation and the influence of PWR environment has caused international concern and triggered numerous research programs [1]. In this context, several codes & standards, including the RCC-M code, have included some technical mandatory or non-mandatory sections to address the issue. In RCC-M, this is compiled in the Rules in Probation Phase 2 and 3 [2]. Due to the novelty of these rules, there is room for improvement for the specific and practical implementation of these rules. AFCEN has hence launched a benchmark exercise at the end of 2019 to help increase the quality of these rules. Part 1 of this paper [3] states that EDF and CNNC/NPIC have launched an effort to benchmark their respective codes on fatigue calculation including the EAF algorithm. In the second part of the benchmark, the two companies started the code comparison based on a benchmark case provided by AFCEN. As stated previously, the 2016 edition of RCC-M code integrates the modifications made to the Code in Probation Phase 2 and 3(RPP)[2], which respectively modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. In this paper, a comparison between RCC-M RPP and NUREG/CR-6909 rev.1 [3] is proposed. The comparison focuses on the technical details of the strain rate calculation and transient combination method. The cumulative fatigue usage factor with or without considering EAF according to RCC-M RPP – 2 and RPP – 3 is given by EDF, using code_aster and its POST_RCCM operator. CNNC/NPIC will provide multiple sets of results including cumulative fatigue usage factors according to RCC-M RPP and NUREG CR/6909 rev. 1 respectively using its own software. Comparison of selection for peak and valleys points, Sn and Fen values are also presented. Differences of the algorithms of the two codes are also discussed.

scholarly journals Helium, [Fe/H] Abundances and the HR (Log TEFF, MBol) Diagram With Hipparcos Data of The Four Nearest Open Clusters: Hyades, Coma Berenices, The Pleiades and Praesepe

1998 ◽  
Vol 11 (1) ◽  
pp. 565-565
Author(s):  
G. Cayrel de Strobel ◽  
R. Cayrel ◽  
Y. Lebreton
Keyword(s):  
Main Sequence ◽  
Open Clusters ◽  
Main Sequence Stars ◽  
Spectroscopic Analyses ◽  
Solar Metallicity ◽  
Metal Abundances ◽  
The Mean ◽  
The Moment ◽  
The One ◽  
Best Fit

After having studied in great detail the observational HR diagram (log Teff, Mbol) composed by 40 main sequence stars of the Hyades (Perryman et al.,1997, A&A., in press), we have tried to apply the same method to the observational main sequences of the three next nearest open clusters: Coma Berenices, the Pleiades, and Praesepe. This method consists in comparing the observational main sequence of the clusters with a grid of theoretical ZAMSs. The stars composing the observational main sequences had to have reliable absolute bolometric magnitudes, coming all from individual Hipparcos parallaxes, precise bolometric corrections, effective temperatures and metal abundances from high resolution detailed spectroscopic analyses. If we assume, following the work by Fernandez et al. (1996, A&A,311,127), that the mixing-lenth parameter is solar, the position of a theoretical ZAMS, in the (log Teff, Mbol) plane, computed with given input physics, only depends on two free parameters: the He content Y by mass, and the metallicity Z by mass. If effective temperature and metallicity of the constituting stars of the 4 clusters are previously known by means of detailed analyses, one can deduce their helium abundances by means of an appropriate grid of theoretical ZAMS’s. The comparison between the empirical (log Teff, Mbol) main sequence of the Hyades and the computed ZAMS corresponding to the observed metallicity Z of the Hyades (Z= 0.0240 ± 0.0085) gives a He abundance for the Hyades, Y= 0.26 ± 0.02. Our interpretation, concerning the observational position of the main sequence of the three nearest clusters after the Hyades, is still under way and appears to be greatly more difficult than for the Hyades. For the moment we can say that: ‒ The 15 dwarfs analysed in detailed in Coma have a solar metallicity: [Fe/H] = -0.05 ± 0.06. However, their observational main sequence fit better with the Hyades ZAMS. ‒ The mean metallicity of 13 Pleiades dwarfs analysed in detail is solar. A metal deficient and He normal ZAMS would fit better. But, a warning for absorption in the Pleiades has to be recalled. ‒ The upper main sequence of Praesepe, (the more distant cluster: 180 pc) composed by 11 stars, analysed in detail, is the one which has the best fit with the Hyades ZAMS. The deduced ‘turnoff age’ of the cluster is slightly higher than that of the Hyades: 0.8 Gyr instead of 0.63 Gyr.

scholarly journals Demonstration of Fatigue for LTO License of NPP Borssele

2015 ◽  
Author(s):  
M. H. C. Hannink ◽  
F. J. Blom ◽  
P. W. B. Quist ◽  
A. E. de Jong ◽  
W. Besuijen
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Integrated Approach ◽  
Temperature Measurements ◽  
Fatigue Assessment ◽  
Term Operation ◽  
Load Monitoring ◽  
Number Of Cycles ◽  
Ageing Management

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands. The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

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