Evaluation of Fretting Fatigue Life for Steam Generator Tubes in Nuclear Power Plants

2007 ◽  
Vol 345-346 ◽  
pp. 243-246 ◽  
Author(s):  
Dae Kyu Park ◽  
Seung Wan Woo ◽  
Il Sup Chung ◽  
Young Suck Chai ◽  
Jae Do Kwon
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fretting Fatigue ◽  
Alloy 600 ◽  
Inconel Alloy ◽  
Alloy 800 ◽  
Incoloy Alloy ◽  
Cyclic Damage

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. INCONEL alloy 600, 690 and INCOLOY alloy 800 are iron-nickel-chromium alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. These alloy are used extensively in the nuclear power plants industry, the chemical industry, the heat-treating industry and the electronic industry. In this paper, the effect of fretting damage on fatigue behavior for INCONEL alloy 600, 690 and INCOLOY alloy 800 were studied. Also, various kinds of mechanical tests such as tension and plain fatigue tests are performed. The objective of this study is to guarantee reliability of the facility applied in a power plant by comparing three materials in respect to fretting fatigue.

The Evaluation of Fretting Fatigue Behavior in Room Temperature for INCOLOY Alloy 800

2007 ◽  
Vol 353-358 ◽  
pp. 89-93 ◽  
Author(s):  
Dae Kyu Park ◽  
Seung Wan Woo ◽  
Yong Tak Bae ◽  
Il Sup Chung ◽  
Young Suck Chai ◽  
...  
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fatigue Behavior ◽  
Fretting Fatigue ◽  
Fatigue Tests ◽  
Chromium Alloy ◽  
Alloy 800 ◽  
Incoloy Alloy

Mechanical breakdown often comes from the fatigue in many structural parts and nuclear power plants. Among the fatigue phenomenon, especially fretting fatigue occurs in mechanical joints showing small relative movements between contact surfaces. Although the research was developed for one hundred years, occurrence mechanism is not clearly identified yet. INCOLOY alloy 800 is a iron-nickel-chromium alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. This alloy is used extensively in the nuclear power plants industry, the chemical industry, the heat-treating industry and the electronic industry. In this paper, the effect of fretting damage on fatigue behavior for INCOLOY alloy 800 was studied. Also, various kinds of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests were carried out with flat-flat contact configuration using a bridge type contact pad and plate type specimen. Through these experiments, it is found that the fretting fatigue strength decreased about 50% compared to the plain fatigue strength. In fretting fatigue, the oblique micro-cracks at an earlier stage are initiated. These results can be used as basic data in a structural integrity evaluation of heat and corrosion resisting alloy considering fretting damages.

A STUDY ON THE FRETTING FATIGUE LIFE OF ZIRCALOY ALLOYS

2008 ◽  
Vol 22 (11) ◽  
pp. 851-856
Author(s):  
JAE-DO KWON ◽  
DAE-KYU PARK ◽  
SEUNG-WAN WOO ◽  
YOUNG-SUCK CHAI
Keyword(s):  
Fatigue Life ◽  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Fatigue Behavior ◽  
Fretting Fatigue ◽  
Mechanical Tests ◽  
Fatigue Tests ◽  
Cyclic Damage

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr , and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.

Analysis of Fatigue Life for Main Steam Isolation Valve in Nuclear Power Plants

2013 ◽  
Author(s):  
Xian Zeng ◽  
Hong Xia ◽  
YeMin Dong ◽  
Zhen Liu
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fatigue Cracks ◽  
Surface Damage ◽  
Fretting Fatigue ◽  
Critical Plane Approach ◽  
Contact Surfaces ◽  
Main Steam

This paper focuses on the analysis of fatigue life for main steam isolation valve (MSIV) in nuclear power plants (NPPs). The contraction of throat is also in the form of sonic resonance cavity, which causes big vibration and noise. The vibration result in the fretting fatigue, which is the reason of the contact surface damage of the disc and the disc guide. By using FEM, the distribution of the stress on the contact surfaces of the disc and the disc guide was calculated. The influences of vibration on the distribution of the stress on the contact surfaces, together with the effect of the contact stress on the initiation of fretting fatigue cracks and fretting fatigue life have been studied. A critical plane approach the Smith–Watson–Topper fatigue parameter has also been adopted here, to predict crack initiation position and orientation, fatigue life. The results show that the predicted cracks location and fatigue life are well accordance with the practical situation.

SCC Inhibitors for SG Tube Materials in Nuclear Power Plants

2007 ◽  
Vol 534-536 ◽  
pp. 717-720
Author(s):  
Kyung Mo Kim ◽  
Eun Hee Lee ◽  
Uh Chul Kim
Keyword(s):  
Electrochemical Properties ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Alloy 600 ◽  
Alloy 800 ◽  
Naoh Solution ◽  
Tubing Material

Several chemicals were studied to suppress the damage due to stress corrosion cracking (SCC) of steam generator (SG) tubes in nuclear power plants. The polarization curves showed that the electrochemical properties on the surface of Alloy 600 MA changed with the addition of inhibitors. The SCC tests were conducted by using a m-RUB specimen in a 10% NaOH solution at a temperature of 315°C. The effects on the SCC of the compounds, TiO2, TyzorLA and CeB6, were tested for several types of SG tubing materials. The test with the addition of TiO2 (P25) and CeB6 showed an effect in decreasing the SCC for the SG tubing material. However, CeB6 caused some more SCC for Alloy 800. The penetration property into a crevice of the inhibitors was investigated by using Alloy 600 specimens with different gap sizes and an AES analysis was performed on the oxide layer of the specimen.

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.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

Experimental Research and Assessment for Thermal Fatigue in T-Junction With Mixing Flows

2017 ◽  
Author(s):  
Chen Yinqiang ◽  
Zhao Xinglei ◽  
Gan Guohua ◽  
Xu Kaili ◽  
Wang Yanzhi ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Stress Distribution ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pipe Wall ◽  
Wall Surface ◽  
Penetration Length ◽  
Test Loop

Thermal fatigue in T-junction with mixing flows has been a grown concern in the wake of the accidents occurring at several nuclear power plants in the world. In this paper, based on highly frequent temperature measurements at the outer pipe wall surface and within the pipe wall from the specially-designed test loop, projected models of turbulence penetration length and temperature distribution were developed. Meanwhile, determining temperature and stress distribution in pipe is the key factor for assessment of thermal fatigue life resulting from hot and cold mixing flows in T-junction. In the paper, the temperature of the inner wall surface was obtained using inverse heat conduction and large eddy simulation based on measurements. Finally, the stress distribution caused by mixing flows was obtained using simplified calculation methods, which were verified by detailed analysis results from the special developed program of inverse function. Such solution methods provide a feasible approach to assess the fatigue life of T-junction in nuclear power plants.

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