Statistical Model of Micro Crack Growth for the Evaluation of Accumulated Fatigue in NPPs

2014 ◽  
Author(s):  
Shigeki Abe ◽  
Takao Nakamura
Keyword(s):  
Statistical Model ◽  
Crack Growth ◽  
Fatigue Failure ◽  
Structural Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
High Reliability ◽  
Reduction Factor ◽  
Cycle Fatigue

In order to establish sophisticated management of aging degradation and to achieve high reliability of components in nuclear power plants (NPPs), it is required to reveal the mechanism of aging degradation and to quantify its deterioration. Present design code requires the assessment of CUF (cumulated usage factor) using design fatigue curve to prevent the occurrence of low-cycle fatigue failure in class 1 components of NPPs. This assessment of CUF prevents the low-cycle fatigue failure effectively up to the present date. However, the margins of structural factor (safety factor) and environmental fatigue reduction factor Fen need to be clarified for the quantitative assessment of aging degradation based on the mechanism of fatigue accumulation in NPPs,. Structural factor and environmental factor are defined in NUREG/CR6909, etc, but they do not clearly explain the technical basis of these factors according to the mechanism of fatigue and environmental effect. In this study, we quantify the dispersions of crack initiation and crack growth in fatigue test in a certain condition aiming at more sophisticated management of aging degradation based on the mechanism of crack growth. Applying these dispersions, we establish a statistical model of micro crack growth to predict fatigue life. We will propose the application of this prediction model of fatigue crack growth to the management of actual components degradation and establish proper maintenance program in order to achieve high reliability of components in NPPs in future.

Initiation and Growth of Small Fatigue Cracks of Steels Used for Nuclear Power Plants Under Low Cycle Regime

2014 ◽  
Author(s):  
Shota Hasunuma ◽  
Takeshi Ogawa
Keyword(s):  
Fatigue Crack ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Fatigue Cracks ◽  
Cycle Fatigue

Low cycle fatigue tests were conducted for carbon steel, STS410, low alloy steel, SFVQ1A, and austenitic stainless steel, SUS316NG, which were used for nuclear power plants, in order to investigate the mechanism of fatigue damage when the plants were subjected to huge seismic loads. In these tests, the surface behavior of fatigue crack initiation and growth was observed in detail using cellulose acetate replicas, while the interior behavior was detected in terms of fracture surface morphology developed by multiple two-step strain amplitude variations with periodical surface removals. Fatigue crack growth rates were evaluated by elasto-plastic fracture mechanics approach. For SFVQ1A and SUS316NG, the fracture mechanics approach is available in order to predict the crack growth life from the metallurgical crack initiation size to the final crack length of the specimens. For STS410, numerous small cracks initiated, grew and coalesced each other on the specimen surface under low cycle fatigue regime.

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.

scholarly journals Evaluation of the Limit State of a Six-Inch Carbon Steel Pipe Elbow in Base-Isolated Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 8400
Author(s):  
Sung-Wan Kim ◽  
Da-Woon Yun ◽  
Bub-Gyu Jeon ◽  
Dae-Gi Hahm ◽  
Min-Kyu Kim
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Base Isolation ◽  
Low Cycle Fatigue ◽  
Limit State ◽  
Steel Pipe ◽  
Cycle Fatigue ◽  
Pipe Elbow ◽  
Isolation Systems

The installation of base isolation systems in nuclear power plants can improve their safety from seismic loads. However, nuclear power plants with base isolation systems experience greater displacement as they handle seismic loads. The increase in relative displacement is caused by the installed base isolation systems, which increase the seismic risk of the interface piping system. It was found that the failure mode of the interface piping system was low-cycle fatigue failure accompanied by ratcheting, and the fittings (elbows and tees) failed due to the concentration of nonlinear behavior. Therefore, in this study, the limit state was defined as leakage, and an in-plane cyclic loading test was conducted in order to quantitatively express the failure criteria for the SCH40 6-inch carbon steel pipe elbow due to low-cycle fatigue failure. The leakage line and low-cycle fatigue curves of the SCH40 6-inch carbon steel pipe elbow were presented based on the test results. In addition, the limit state was quantitatively expressed using the damage index, based on the combination of ductility and energy dissipation. The average values of the damage index for the 6-inch pipe elbow calculated using the force−displacement (P–D) and moment−relative deformation angle (M–R) relationships were found to be 10.91 and 11.27, respectively.

scholarly journals Parameters influencing the low-cycle fatigue life of materials in pressure water reactor nuclear power plants / Parametry ovlivòující únavové chování materiálù pro tlakovodní jaderné reaktory v režimu nízkocyklové únavy

2015 ◽  
Vol 59 (3) ◽  
pp. 91-98
Author(s):  
V. Šefl
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Carbon Steel ◽  
Austenitic Stainless Steel ◽  
Literature Review ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue

Abstract In this literature review we identify and quantify the parameters influencing the low-cycle fatigue life of materials commonly used in nuclear power plants. The parameters are divided into several groups and individually described. The main groups are material properties, mode of cycling and environment parameters. The groups are further divided by the material type - some parameters influence only certain kind of material, e.g. sulfur content may decreases fatigue life of carbon steel, but is not relevant for austenitic stainless steel; austenitic stainless steel is more sensitive to concentration of dissolved oxygen in the environment compared to the carbon steel. The combination of parameters i.e. conjoint action of several detrimental parameters is discussed. It is also noted that for certain parameters to decrease fatigue life, it is necessary for other parameter to reach certain threshold value. Two different approaches have been suggested in literature to describe this complex problem - the Fen factor and development of new design fatigue curves. The threshold values and examples of commonly used relationships for calculation of fatigue lives are included. This work is valuable because it provides the reader with long-term literature review with focus on real effect of environmental parameters on fatigue life of nuclear power plant materials.

1715 Behavior of AE generation during low cycle fatigue test of steels for nuclear power plants

2010 ◽  
Vol 2010 (0) ◽  
pp. 744-745
Author(s):  
Makoto OHTA ◽  
Yoshihiro MIZUTANI ◽  
Akira TODOROKI ◽  
Ryosuke MATSUZAKI
Keyword(s):  
Fatigue Test ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue

Predicting Catastrophic OD Initiated Fatigue Failure of Thick-Walled Cylinders Using Low Cycle Fatigue Data

1983 ◽  
Vol 22 ◽  
Author(s):  
Joseph A. Kapp
Keyword(s):  
Crack Growth ◽  
Fatigue Failure ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue ◽  
Analysis Method ◽  
Life Data ◽  
Fatigue Data ◽  
Concentration Factors ◽  
Crack Initiation Life ◽  
Stress Concentration Factors

ABSTRACTA procedure is presented to predict fatigue failure of thick-walled cylinders containing discontinuities at the OD. Both crack initiation life and crack growth are considered. The elastic-plastic strains at an OD discontinuity are estimated using an elastic analysis and stress concentration factors. The strain estimates are then used in conjunction with low cycle fatigue data to determine the initiation life. Crack growth life is estimated by integration of a power law relationship. The results obtained by using this analysis method compared to measured fatigue life data for several OD initiated failures in thick-walled cylinders agrees to within about 10 percent.

Low Cycle Fatigue: Probability and Statistical Modeling of Fatigue Life

2014 ◽  
Author(s):  
D. Gary Harlow
Keyword(s):  
Distribution Function ◽  
Fatigue Life ◽  
Damage Accumulation ◽  
Power Plants ◽  
Structural Reliability ◽  
Low Cycle Fatigue ◽  
High Reliability ◽  
Cycle Fatigue ◽  
Low Carbon ◽  
Side Body

Low cycle fatigue (LCF) induces damage accumulation in structural components used in various applications. LCF typically describes conditions for which plastic strains are larger than elastic strains. In order to certify and qualify a structural component, manufactured from a given material, that requires high reliability for operation and safety, fundamental material properties should be experimentally investigated and validated. The traditional strain–life approach serves as the underlying experimental method for most LCF investigations. Building upon that background, the purpose of this paper is to investigate the statistical variability and appropriately model that variability for life in LCF. Specifically, the variability associated with the median behavior in a strain–life graph for data is examined. The ensuing analyses are based on data for a cold-rolled, low carbon, extra deep drawing steel; ASTM A969 which is appropriate for applications where extremely severe drawing or forming is envisioned. It is frequently used in the automotive industry for components such as inner door components and side body components. For substantiation of the proposed modeling techniques, data for 9Cr-1Mo steel is also investigated. Such steel is frequently used in the construction of power plants and other structures that experience operating temperatures in excess of 500°C. The commonly used universal slopes approach for fatigue life modeling for which the strain–life computation employs the standard Coffin–Manson relationship is compared to a statistical methodology using a distribution function frequently used in structural reliability. The proposed distribution function for characterizing the fatigue life is a generalized Weibull distribution function that empirically incorporates load history and damage accumulation.

Complex Low-Cycle Fatigue Damage Assessment of Components and Piping of Nuclear Power Plants Type WWER

2017 ◽  
Author(s):  
Libor Vlcek ◽  
Lubomir Junek
Keyword(s):  
Fatigue Damage ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Life Assessment ◽  
Design Stage ◽  
Multiaxial Loading ◽  
Cycle Fatigue ◽  
Mathematical Formulas

An innovative principle of low-cycle fatigue (LCF) life assessment suggested for WWER nuclear power plants is presented. In the design stage the fatigue life assessment is based on fatigue design curves, which are introduced in graphical form for air environment. Alternatively and especially for operational stage the fatigue curves are constructed on the basis of mathematical formulas. Mathematical descriptions were validated by strain-controlled LCF laboratory tests. Due to such validated mathematical formulas the complex LCF damage analyses of nuclear power plant components and piping are enabled. In the frame of complex LCF assessment the influence of operating temperatures, stress asymmetry ratio, corrosion environment, neutron fluency and multiaxial loading can be taken into account not only for the base steel materials, but also for their welds. The aim of this paper is to summarise the whole methodology of complex LCF assessment and damage prediction including operational limits of fatigue damage defined in the Czech nuclear standard. The innovation process of original Russian LCF formulas has been running since 2010 based on three national R&D projects focused mainly on environmental aspects and multiaxial loading.

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