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.

Experimental Fretting-Wear Studies of Steam Generator Materials

1995 ◽  
Vol 117 (4) ◽  
pp. 312-320 ◽  
Author(s):  
N. J. Fisher ◽  
A. B. Chow ◽  
M. K. Weckwerth
Keyword(s):  
Steam Generator ◽  
Contact Force ◽  
Environmental Conditions ◽  
Analytical Techniques ◽  
Work Rate ◽  
Fretting Wear ◽  
Flow Induced Vibration ◽  
Dynamic Interactions ◽  
Support Materials ◽  
Wear Damage

Flow-induced vibration of steam generator tubes results in fretting-wear damage due to impacting and rubbing of the tubes against their supports. This damage can be predicted by computing tube response to flow-induced excitation forces using analytical techniques, and then relating this response to resultant wear damage using experimentally derived wear coefficients. Fretting-wear of steam generator materials has been studied experimentally at Chalk River Laboratories for two decades. Tests are conducted in machines that simulate steam generator environmental conditions and tube-to-support dynamic interactions. Different tube and support materials, tube-to-support clearances, and tube support geometries have been studied. The effect of environmental conditions, such as temperature, oxygen content, pH and chemistry control additive, have been investigated as well. Early studies showed that damage was related to contact force as long as other parameters, such as geometry and motion, were held constant. Later studies have shown that damage is related to a parameter called work-rate, which combines both contact force and sliding distance. Results of short and long-term fretting-wear tests for CANDU steam generator materials at realistic environmental conditions are presented. These results demonstrate that work-rate is an appropriate correlating parameter for impact-sliding interaction.

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.

Heat Exchanger Tube Fretting Wear: Review and Application to Design

1985 ◽  
Vol 107 (2) ◽  
pp. 149-156 ◽  
Author(s):  
P. L. Ko
Keyword(s):  
Heat Exchanger ◽  
Impact Force ◽  
Design Criterion ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Dynamic Interactions ◽  
Tube Life ◽  
The Impact

Flow-induced vibration in steam generators and heat exchangers can cause dynamic interactions between tubes and tube supports resulting in fretting-wear. The effects on tube wear of various parameters, such as tube/support interactions, materials, and tube/support clearances have been studied. Techniques to predict the dynamic tube/support interaction and analyze the impact force at the support have been developed. The results of this work are reviewed and discussed in the context of how best they may be applied in the assessment of heat exchanger designs. A new design criterion based on support impact force is also discussed. Finally, a technique to predict long-term tube life is outlined.

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.

Vibration Excitation Forces in a Rotated Triangular Tube Bundle Subjected to Two-Phase Cross Flow

2010 ◽  
Author(s):  
H. Senez ◽  
N. W. Mureithi ◽  
M. J. Pettigrew
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Cross Flow ◽  
Fretting Wear ◽  
Experimental Program ◽  
Phase Flow ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Fluid Forces ◽  
Vibration Excitation

Two-phase cross flow exists in many shell-and-tube heat exchangers. Flow-induced vibration excitation forces can cause tube motion that will result in long-term fretting wear or fatigue. Detailed flow and vibration excitation force measurements in tube bundles subjected to two-phase cross flow are required to understand the underlying vibration excitation mechanisms. Studies on this subject have already been done, providing results on flow regimes, fluidelastic instabilities, and turbulence-induced vibration. The spectrum of turbulence-induced forces has usually been expected to be similar to that in single-phase flow. However, a recent study, using tubes with a diameter larger than that in a real steam generator, showed the existence of significant quasi-periodic forces in two-phase flow. An experimental program was undertaken with a rotated-triangular array of cylinders subjected to air-water cross-flow, to simulate two-phase mixtures. The tube bundle here has the same geometry as that of a real steam generator. The quasi-periodic forces have now also been observed in this tube bundle. The present work aims to understand turbulence-induced forces acting on the tube bundle, providing results on drag and lift force spectra and their behaviour according to flow parameters, and describing their correlations. Detailed experimental test results are presented in this paper. Comparison is also made with previous measurements with larger diameter tubes. The present results suggest that quasi-periodic fluid forces are not uncommon in tube arrays subjected to two-phase cross-flow.

Experimental Study of Dynamic Interaction Between a Steam Generator Tube and an Anti-Vibration Bar

2010 ◽  
Author(s):  
V. Lalonde ◽  
A. Ross ◽  
M. J. Pettigrew ◽  
I. Nowlan
Keyword(s):  
Experimental Study ◽  
Dynamic Behavior ◽  
Steam Generator ◽  
Contact Force ◽  
Dynamic Interaction ◽  
Work Rate ◽  
Fretting Wear ◽  
Steam Generators ◽  
Simply Supported ◽  
Excitation Force

A first experimental work was previously carried out to study the dynamic behavior of a tube simply supported at both ends in interaction with an anti-vibration bar at mid-span. This paper presents modifications to the previous setup with the aim of improving the accuracy of the results. A comparison of the dynamic behavior of the tube is made between both setups. The objective of this experimental study is to characterize the vibration behavior of U-tubes found in steam generators of nuclear power plants. Indeed, two-phase cross-flow in the U-tubes section of steam generators can cause many problems related to vibration. In fact, flow-induced vibration of the U-tubes can cause impacts or rubbing of the tubes against their flat bar supports. Variation of the clearance between the AVB and the U-tubes may lead to ineffective supports. The resulting in-plane and out-of-plane motions of the tubes are causing fretting-wear and impact abrasion. In this study, the clearance between the tube and the AVB, as well as the amplitude, form and frequency of the excitation force are controlled parameters. The first two modes of the tube are studied. The modifications made to the setup lead to significant improvements in the results. The natural frequencies of both setups are compared to theoretical values. The difference between experimental and theoretical frequencies confirms that the new setup better represents the theoretical model of a simply supported tube. The damping of both setups is also compared to values found in literature. The results show that the new setup is more representative of realistic steam generator situations. Compared to the first setup, the displacements of the new setup clearly indicate that the movement of the tube is mostly parallel to the flat bar and in the same direction as the excitation force. The whirling motion of the tube is prevented in the new setup. The accuracy of the contact force as a function of clearance was also improved. The use of more sensitive force sensors helped to reduce the noise level of the contact force. Finally, the dynamic interaction between the tube and the AVB, defined by the fretting wear work-rate, presents a more consistent behavior. The maximum work-rate occurs when the tube is excited around the second mode for clearance between −0.10 and 0.00 mm. Such clearance between the tube and the AVB should then be avoided to minimize fretting damage.

Degradation Mechanism of Dissimilar Metal Weld Joints on Steam Generator Collectors VVER 440MW Type

2017 ◽  
Vol 741 ◽  
pp. 134-137
Author(s):  
Lubomír Junek ◽  
Ladislav Jurasek ◽  
Zdeněk Čančura ◽  
Miroslava Ernestová ◽  
Zuzana Skoumalová
Keyword(s):  
Steam Generator ◽  
Experimental Analysis ◽  
Degradation Mechanism ◽  
Steam Generators ◽  
Dissimilar Metal ◽  
Weld Joints ◽  
Dissimilar Metal Weld ◽  
Metal Weld

Indications were detected on dissimilar metal welds (DMW) of steam generators (SG) after 20 years of operation during NDT inspections. Indications started slowly growth every year. DMW on SGs had to be repaired. Paper describes experimental analysis and degradation mechanism of SG weld joints failures.

Steam Generator In-Service Inspections at Paks NPP

2008 ◽  
Author(s):  
Miklo´s Do´czi
Keyword(s):  
Heat Exchanger ◽  
Steam Generator ◽  
Eddy Current ◽  
Power Plants ◽  
Tube Bundle ◽  
Degradation Mechanism ◽  
Feed Water ◽  
Material Damage ◽  
Steam Generators ◽  
Service Inspection

Steam Generator is one of the most critical components in nuclear power plants. It has of overriding importance from point of view of safe and reliable operation of the whole plant. Variety of degradation mechanisms affecting SG tube bundle may cause different types of material damage. In Paks NPP eddy current in-service inspection have been performed since 1988. In the year 1997 higher number of defected tubes were found in case of Unit#2, compared to results of the previous years. A medium term SG inspection program had been performed in the time period between 1998–2004. Based on the results of eddy current inspections high number of heat exchanger tubes had been plugged. Chemical cleanings of all steam generators were performed aiming to reduce the magnetite, copper deposits and corrosion agents acting on the surface of the tubes. Replacement of the main condensers had been performed to stop the uncontrolled water income caused by the relatively frequent leakages of the condenser tubes. Several tube samples had been cut from the first row of the tube bundles of different steam generators to study the effectiveness of the cleaning process and to determine the composition of deposits on the tube outside surface. Also several tubes with eddy current indications had been pulled out from the steam generators to determine the acting degradation mechanism. Examination of removed tubes can provide opportunity to check the reliability of eddy current inspection using bobbin coil. Also there were tubes pulled out form SG with existing cracks. From the year 2005 new inspection program had been started. As the first results of the new inspection program shows, there is only a few new indications had been found and there is no measurable crack propagation in case of existing indications. During the recent years feed-water collectors were replaced in case of all units of the power plant, because of material damage (erosion corrosion). The paper summarizes the results of eddy current in-service inspection of heat exchanger tubes, results of examinations of removed tubes and also deals with results of visual examination of the feed-water distributor system.

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.

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