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.

Failure criteria of a carbon steel pipe elbow for low-cycle fatigue using the damage index

2020 ◽  
Vol 153 ◽  
pp. 106800
Author(s):  
Sung-Wan Kim ◽  
Sung-Jin Chang ◽  
Dong-Uk Park ◽  
Bub-Gyu Jeon
Keyword(s):  
Carbon Steel ◽  
Low Cycle Fatigue ◽  
Damage Index ◽  
Failure Criteria ◽  
Steel Pipe ◽  
Cycle Fatigue ◽  
Pipe Elbow

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.

Strain Rate Effects in SA-106 Carbon Steel Pipe

1982 ◽  
Vol 104 (1) ◽  
pp. 31-35 ◽  
Author(s):  
D. Peterson ◽  
J. E. Schwabe ◽  
D. G. Fertis
Keyword(s):  
Carbon Steel ◽  
Yield Strength ◽  
Strain Rate ◽  
Tensile Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Experimental Modeling ◽  
Steel Pipe ◽  
Strain Rate Effects ◽  
Rate Effects

Experiments were performed to measure the effect of strain rate on the tensile properties of SA-106 carbon steel pipe, in support of analysis and experimental modeling of postulated pipe whip in nuclear power plants. It was observed that increasing the strain rate from 4 × 10−4 to 4 s−1 raised the yield strength by approximately 30 percent.

Dynamic Testing of High Density Polyethylene Vent and Drain Configurations

2012 ◽  
Author(s):  
Douglas Munson ◽  
Timothy M. Adams ◽  
Shawn Nickholds
Keyword(s):  
Carbon Steel ◽  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Testing ◽  
High Density ◽  
Steel Pipe ◽  
Labor Costs ◽  
Service Water ◽  
Hdpe Pipe

For corroded piping in low temperature systems, such as service water systems in nuclear power plants, replacement of carbon steel pipe with high density polyethylene (HDPE) pipe is a cost-effective solution. HDPE pipe can be installed at much lower labor costs than carbon steel pipe, and HDPE pipe has a much greater resistance to corrosion. This paper presents the results of the seismic testing of selected vent and drain configurations. This testing was conducted to provide proof of the conceptual design of HDPE vent and drain valve configurations. A total of eight representative models of HDPE vent and drain assemblies were designed. The models were subjected to seismic SQURTS spectral acceleration up to maximum shake table limits. The test configurations were then checked for leakage and operability of the valves. The results for these tests, along with the test configurations, are presented. Also presented are the acceleration data observed at various points on the test specimens.

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

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.

scholarly journals Experimental Stress Analysis for Four 24-in. ANSI Standard B16.9 Tees

1977 ◽  
Vol 99 (4) ◽  
pp. 537-552 ◽  
Author(s):  
J. K. Hayes ◽  
S. E. Moore
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Low Cycle Fatigue ◽  
Branch Pipe ◽  
Fatigue Tests ◽  
Elastic Response ◽  
Experimental Stress ◽  
Stress Distributions ◽  
Cycle Fatigue ◽  
Experimental Stress Analysis

This paper describes the experimental stress analysis of low cycle fatigue tests of four tees tested by Combustion Engineering, Inc. (C-E) under subcontract to Union Carbide Nuclear Division. These tests are part of the ORNL Design Criteria for Piping and Nozzles Program which is being conducted for the development of design criteria for nuclear power plant service piping components. The test assemblies were fabricated at C-E from commercially obtained ANSI B16.9 tees and matching diameter steel pipes welded to the tees, with suitable end closures and fixtures for applying the loads. The tees tested and discussed in this report are described in the following: Tee Number/Material/Nominal Size: T–11/carbon steel/24″×24″×24″ sch 160; T–12/carbon steel/24″×24″×10″ sch 40; T–13/carbon steel/24″×24″×10″ sch 160; T–16/stainless steel/24″×24″×24″ sch 10. Each tee test assembly was instrumented with approximately 240 rectangular strain gage rosettes for determining elastic stress distributions, and six linear variable displacement transducers for determining flexibility factors. Elastic-response tests were conducted for 12 loading conditions consisting of internal pressure, pure bending and torsional moments and direct force loads applied individually to the branch pipe extension and to one end of the run pipe. The other run pipe extension was fixed rigidly to the loading frame. Automatic data handling equipment and data reduction techniques were used to process the strain gage readings. For each loading condition, stress distributions were determined and the locations and magnitudes of the maximum stresses were identified. Test results are presented and compared with appropriate design formulas of the ASME Boiler and Pressure Vessel Code, Section III. After the elastic-response tests were completed, three of the tees were low-cycle fatigue tested by pressure cycling using transformer oil. The T-11 and T-13 tees were pressure cycled for 100 psig (790 kPa) to 7000 psig (48 360 kPa); whereas, the T-12 tee was pressure cycled from 0 psig (100 kPa) to 1800 psig (12 510 kPa). A low-cycle fatigue test was performed on the T-16 tee assembly by applying a bending moment to the branch pipe in the plane of the tee with the tee pressurized to a constant internal pressure of 300 psig (2170 kPa). All low-cycle fatigue tests were performed until a through-the-wall fatigue crack occurred as evidenced by a leak. Subparagraph NB-3653.6 of ASME Code, Section III, Division I, Nuclear Power Plant Components was used to calculate the fatigue design life and comparisons were made with the experimentally determined fatigue life.

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