Wear Characteristics of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions

2007 ◽  
Vol 120 ◽  
pp. 181-186 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Young Ze Lee
Keyword(s):  
Wear Mechanisms ◽  
Nuclear Power ◽  
Fluid Flows ◽  
Fretting Wear ◽  
Wear Volume ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Characteristics ◽  
Contact Surfaces ◽  
Contact Part

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature flows in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration) which cause fretting wear in contact part of tube-support. The reduction of tube thickness due to fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research on the fretting wear characteristics of tube-support is required. This work is focused on investigations of fretting wear characteristics and wear mechanisms of tube-support. Results show that the wear rate of tube is proportional to that of support and that with increasing the water temperature the wear volume of tube-support decreases because the oxidation rate decreases due to reduction of the oxygen concentration in contact surfaces. Also, the wear mechanisms of tube-support are abrasive and oxidational wear.

Fretting Wear Characteristics of Tube-Support Components for a Steam Generator in Elevated Temperature

2005 ◽  
Author(s):  
Sung-Hoon Jeong ◽  
Young-Ze Lee
Keyword(s):  
High Pressure ◽  
Wear Mechanisms ◽  
Nuclear Power ◽  
Fretting Wear ◽  
Wear Volume ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Characteristics ◽  
Contact Surfaces ◽  
Contact Part

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid of high-pressure and high-temperature flows in the tubes and the flows cause oscillating motions between tubes and supports. This is called FIV (flow induced vibration) which cause fretting wear in contact part of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on investigations of fretting wear characteristics and wear mechanisms of tube-support. Results are that the wear rate of tube is proportional to that of support and as water temperature increases the wear volume of tube-support decreases because the oxidation rate decreases due to lack of the oxygen concentration in contact surfaces. Also, the wear mechanisms of tube-support are abrasive and oxidational wear.

Wear Transitions of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions

2006 ◽  
Vol 326-328 ◽  
pp. 1263-1266 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Jung Min Park ◽  
Joong Hui Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Applied Load ◽  
Fluid Flows ◽  
Relative Displacement ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Support Materials ◽  
Mild Wear ◽  
Inconel 690

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on fretting wear transitions from mild wear to severe wear of tube-support materials by various loads and relative displacements. The transition is defined on the basis of the changes in wear amount. To investigate the transition, the fretting wear tester was contrived to prevent the reduction of relative displacement between tube and support by increasing the load. The tube and support materials were Inconel 690 and 409 SS, respectively. The results show that the transition of tube-support wear is caused by the changes of the dominant wear mechanism depending on the applied load and the relative displacement.

Wear Transitions of Tube-Support Materials for a Nuclear Steam Generator through Sliding Wear Test and Fretting Wear Test

2006 ◽  
Vol 321-323 ◽  
pp. 430-433 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Byoung Jong Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Fluid Flows ◽  
Wear Test ◽  
Relative Displacement ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Support Materials ◽  
Mild Wear ◽  
Inconel 690

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on fretting wear transitions from mild wear to severe wear of tube-support materials by various loads and relative displacements. The transition is defined on the basis of the changes in wear amount. To investigate the transition, the fretting wear tester was contrived to prevent the reduction of relative displacement between tube and support by increasing the load. The tube and support materials were Inconel 690 and 409 SS, respectively. The results show that the transition of tube-support wear is caused by the changes of the dominant wear mechanism depending on the applied load and the relative displacement.

Wear Characteristics of INCONEL 690 and INCONEL 600 in Elevated Temperature

2005 ◽  
Vol 297-300 ◽  
pp. 1424-1429 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Young Ze Lee
Keyword(s):  
Wear Mechanisms ◽  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Fretting Wear ◽  
Wear Characteristics ◽  
Inconel 600 ◽  
Wear Life ◽  
Inconel 690 ◽  
C 50

In this paper, the fretting wear characteristics of INCONEL 690 (I-690) and INCONEL 600 (I-600) was evaluated to verify the wear mechanism and the wear life. Because of the excellent corrosion-resistance of nickel-based alloy, those materials are used for steam generator tube in nuclear power plants. Sometimes the tubes are damaged due to small amplitude vibration, so called fretting wear. To verify the fretting wear mechanisms the wear experiment was carried with the crossed-cylinder wear tester, which used a cam to oscillate the specimen. The test was carried out at loads of 40N and 90N in elevated temperatures of water. The temperatures of water were 20°C, 50°C and 80°C. The increase of water temperature causes the oxidation of the contact area to be delayed, and the amount of wear on oxide layer to be reduced. The main wear mechanisms of fretting were abrasive wear and oxidation wear.

Fretting Wear Damage of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Grids

2007 ◽  
Vol 345-346 ◽  
pp. 709-712
Author(s):  
Jin Seon Kim ◽  
Yong Hwan Kim ◽  
Seung Jae Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Fluid Flows ◽  
Fretting Wear ◽  
Wear Depth ◽  
Fuel Cladding ◽  
Flow Induced Vibration ◽  
Nuclear Fuel Assembly ◽  
Wear Damage ◽  
Increasing Load

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

EXPERIMENTAL STUDY ON IMPACT FRETTING WEAR OF INCONEL TUBES UNDER HIGH TEMPERATURE AND PRESSURE

2011 ◽  
Vol 25 (31) ◽  
pp. 4253-4256 ◽  
Author(s):  
CHOON YEOL LEE ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI ◽  
KYOOSIK SHIN
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Impact Force ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Experimental Parameters ◽  
Generator Design ◽  
The Impact

In nuclear power plant, fretting wear caused by flow induced vibration (FIV) accompanied with impact force can make serious problems between U -tubes and egg-crates which are located in steam generators. In order to guarantee the reliability of the steam generator, design based on consideration of the damage due to the fretting wear of the U -tube is inevitable. The purpose of this study is to elucidate fretting wear mechanism qualitatively and quantitatively. First, finite element models are developed to analyze the dynamic characteristics and estimate the impact force in steam generators. Based on the numerical results, fretting wear simulation is performed according to the environment to which the actual steam generators in nuclear power plant are exposed. Initial experimental results are obtained for various experimental parameters and the effect of work rate and temperature on fretting wear is evaluated.

Finite element analysis of fretting wear considering variable coefficient of friction

2018 ◽  
Vol 233 (5) ◽  
pp. 758-768 ◽  
Author(s):  
Ling Li ◽  
Le Kang ◽  
Shiyun Ma ◽  
Zhiqiang Li ◽  
Xiaoguang Ruan ◽  
...  
Keyword(s):  
Finite Element ◽  
Coefficient Of Friction ◽  
Variable Coefficient ◽  
Friction Model ◽  
Slip Condition ◽  
Fretting Wear ◽  
Partial Slip ◽  
Wear Volume ◽  
The Coefficient Of Friction ◽  
Contact Surfaces

Fretting wear is a kind of material damage in contact surfaces caused by microrelative displacement between two bodies. It can change the profile of contact surfaces, resulting in loosening of fasteners or fatigue cracks. Finite element method is an effective method to simulate the evolution of fretting wear process. In most studies of fretting wear, the coefficient of friction was assumed to be constant to simplify model and reduce the difficulty of solving. However, fretting wear test showed that the coefficient of friction was a variable related to the number of fretting cycles. Therefore, this paper introduces the coefficient of friction as a function of the number of fretting cycles in numerical simulation. A wear model considering variable coefficient of friction is established by combining energy consumption model and adaptive grid technique. The nodes of contact surfaces are updated through the UMESHMOTION subroutine. The effects of constant coefficient of friction and variable coefficient of friction on fretting wear are analyzed by comparing the wear amount under different loading conditions. The results show that when compared with coefficient of friction model, fretting wear is obviously affected by variable coefficient of friction and the variable coefficient of friction model has a larger wear volume when the fretting is in partial slip condition and mixed slip condition. In gross slip condition, the difference of wear volume between variable coefficient of friction model and coefficient of friction model decreases with the increase in the displacement amplitudes.

Vibration Analysis of Steam Generators and Heat Exchangers: An Overview — Part II: Vibration, Response, Fretting-Wear, Guidelines

2002 ◽  
Author(s):  
Michel J. Pettigrew ◽  
Colette E. Taylor
Keyword(s):  
Heat Exchangers ◽  
Vibration Analysis ◽  
Response Prediction ◽  
Design Guidelines ◽  
Vibration Response ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Service Water

Design guidelines were developed to prevent tube failures due to excessive flow-induced vibration in shell-and-tube heat exchangers. An overview of vibration analysis procedures and recommended design guidelines is presented in this paper. This paper pertains to liquid, gas and two-phase heat exchangers such as nuclear steam generators, reboilers, coolers, service water heat exchangers, condensers, and moisture-separator-reheaters. Part 2 of this paper covers forced vibration excitation mechanisms, vibration response prediction, resulting damage assessment, and acceptance criteria.

Probabilistic Assessment of Fretting Wear in Steam Generator Tubes Under Flow Induced Vibrations

2003 ◽  
Author(s):  
Greg D. Morandin ◽  
Richard G. Sauve´
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Probabilistic Methods ◽  
Fretting Wear ◽  
Nonlinear Flow ◽  
Impact Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Damage ◽  
The Impact

Successful life management of steam generators requires an ongoing operational assessment plan to monitor and address all potential degradation mechanisms. A degradation mechanism of concern is tube fretting as a result of flow-induced vibration. Flow induced vibration predictive methods routinely used for design purposes are based on deterministic nonlinear structural analysis techniques. In previous work, the authors have proposed the application of probabilistic techniques to better understand and assess the risk associated with operating power generating stations that have aging re-circulating steam generators. Probabilistic methods are better suited to address the variability of the parameters in operating steam generators, e.g., flow regime, support clearances, manufacturing tolerances, tube to support interactions, and material properties. In this work, an application of a Monte Carlo simulation to predict the propensity for fretting wear in an operating re-circulation steam generator is described. Tube wear damage is evaluated under steady-state conditions using two wear damage correlation models based on the tube-to-support impact force time histories and work rates obtained from nonlinear flow induced vibration analyses. Review of the tube motion in the supports and the impact/sliding criterion shows that significant tube damage at the U-bend supports is a result of impact wear. The results of this work provide the upper bound predictions of wear damage in the steam generators. The EPRI wear correlations for sliding wear and impact wear indicate good agreement with the observed damage and, given the preponderance of wear sites subject to impact, should form the basis of future predictions.

Design Specifications to Ensure Flow-Induced Vibration and Fretting-Wear Performance in CANDU™ Steam Generators and Heat Exchangers

2009 ◽  
Author(s):  
Victor Janzen ◽  
Yingke Han ◽  
Michel Pettigrew
Keyword(s):  
Heat Exchangers ◽  
Dynamic Properties ◽  
Fretting Wear ◽  
Performance Criteria ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Design Specifications ◽  
Field Evidence ◽  
Main Steam

Preventing flow-induced vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. This paper outlines the steps required to generate and support such design specifications for CANDU™ nuclear steam generators and heat exchangers, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system “excursions” or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues.

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