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

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.

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.

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.

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.

Structure and Mechanical Properties of Welded Joints for Nuclear Power Plants of Type MIR 1200

2017 ◽  
Vol 891 ◽  
pp. 201-205
Author(s):  
Ladislav Kander ◽  
Petr Čížek ◽  
Šárka Hermanová ◽  
Zdeněk Říha
Keyword(s):  
Mechanical Properties ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Shear Fracture ◽  
Low Cycle Fatigue ◽  
Water Environment ◽  
Temperature Shift ◽  
Pressure Vessels

The paper deals with research, development and verification of production technology of selected welded joints for pressure vessels of primary circuits of nuclear power plants of type MIR 1200. Effect of various welding technology including simulation heat treatment on mechanical and fracture properties have been studied. Four type of homogenous 10GN2MFA – 10GN2MFA type of welded joints have been prepared for experimental programme. Conventional mechanical properties (tensile and impact test) as well as unconventional mechanical properties (fracture mechanics, low-cycle fatigue and stress corrosion cracking in water environment) have been studied. Effect of elevated working temperature on structure and material properties has been evaluated. Temperature dependencies of shear fracture have been plotted and effect of welding procedure on transition temperature shift has been evaluated. Experimental data have been compared with numerical simulation using FEM.

Experimental Stress Analysis of 24-In. Tees

1971 ◽  
Vol 93 (4) ◽  
pp. 919-928 ◽  
Author(s):  
J. K. Hayes ◽  
B. Roberts
Keyword(s):  
Stress Analysis ◽  
Fatigue Test ◽  
Nuclear Power ◽  
Low Cycle Fatigue ◽  
National Laboratory ◽  
Branch Pipe ◽  
Experimental Stress ◽  
Cycle Fatigue ◽  
Oak Ridge ◽  
Experimental Stress Analysis

This paper describes the experimental stress analysis and low cycle fatigue test of one 24-in. dia, schedule 40 carbon steel, ASA Standrd B16.9 tee performed by Combustion Engineering, Inc. This program is part of the Oak Ridge National Laboratory Piping Program for the development of design criteria for nuclear service piping components. The tee was instrumented with 230 rectangular strain gage rosettes. Elastic data was obtained for 12 loading conditions consisting of internal pressure and orthogonal pure moment and orthogonal direct forces applied individually to the free branch and run ends of the tee. One of the run ends of the tee was “built in” throughout the test. All loads were applied through pipe extensions welded to the tee. The tee was tested to failure in a low cycle fatigue test with an in-plane bending moment on the branch pipe. The tee was pressurized to the design pressure of 1025 psig during the fatigue test. A cyclic stress of approximately ±83,600 psi was imposed on the tee and a through-the-wall fatigue crack occurred at 18,532 cycles. Significant test results are summarized and compared with design values tabulated in the current issue of the Nuclear Power Piping Code, USAS B31.7-1969.

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