The Effect of Coating on the Fretting Wear Between Tubes and Supports in a Nuclear Steam Generator

2007 ◽  
Author(s):  
Joong-Hui Lee ◽  
Jin-Sun Kim ◽  
Jung-Min Park ◽  
Bo-Ra Shin ◽  
Young-Ze Lee
Keyword(s):  
Steam Generator ◽  
Coating Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Hard Coatings ◽  
Fretting Wear ◽  
Tube Material ◽  
Flat Type ◽  
Wear Damage ◽  
Wear Tests

The steam generator in the nuclear power plants is a kind of heat exchanger, which is composed of bundles of long and slender pipes. The tubes are supported by anti-vibration bar to reduce the vibration, caused by the water flows for cooling purpose. The wear damage due to this vibration is called as the fretting wear, which should be minimized for the safe operation of plants. The hard coatings are very effective to reduce the amount of wear. In this paper, the coatings of TiN and CrN were deposited on the tube material to protect the fretting surfaces. The tube-on-flat type tester was used for fretting wear tests. The wear amounts of the coated tubes were decreased depending on coating thickness. CrN coating was very effective to reduce the wear. In case of TiN the wear amounts were dependent on the coating thickness. Thick coating of TiN was very effective for wear resistance.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

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.

Development of Wear Test System for Steam Generator Tubes in Nuclear Power Plants

2005 ◽  
Vol 297-300 ◽  
pp. 1418-1423 ◽  
Author(s):  
Chi Yong Park ◽  
Yong Sung Lee ◽  
Myung Hwan Boo
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Wear Test ◽  
Test System ◽  
Fretting Wear ◽  
Stable Operation ◽  
Test Results ◽  
Steam Generator Tubes

In steam generators of nuclear power plants, flow-induced vibration (FIV) can lead to tube damage by fretting-wear occurred due to impact and sliding movement between the tubes and their supports. There have been many studies and test results on wear damage of steam generator tubes but they were not reflected the mechanical and chemical conditions accurately. KEPRI nuclear power laboratory developed a wear test system, which is able to control the motion of impact and sliding simultaneously in the pressurized high temperature water-chemistry conditions. Some wear tests were performed to verify the stable operation for the wear test. This wear test system with new concepts was described briefly, and some data for verifying its performance have been shown in the cases of the selected some test results. In the test, Alloy 690 was used for tube materials and 409 stainless steel for support plates. A little data deviation was obtained and stability of system operation was investigated.

A Numerical Characterization of Flow-Induced Vibration and Fretting Wear Potential in Nuclear Steam Generator Tube Bundles

2011 ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Flow Induced Vibration ◽  
Numerical Characterization ◽  
Wear Damage ◽  
Effect Of Support ◽  
Tube Failure

Nuclear power plants have experienced problems related to tube failures in steam generators. While many of these failures have been attributed to corrosion, it has been recognized that flow-induced vibrations contribute significantly to tube failure. In order to avoid these excessive vibrations, tubes are stiffened by placing supports along their length. Various tube/support geometries have been used, but the majority are either support plates (plates with drilled or broached holes) or flat bars. Unfortunately, clearance is often considered necessary between the tubes and their supports to facilitate tube/support assembly and to allow for thermal expansion of the tubes. A combination of flow-induced turbulence and fluidelastic forces may then lead to unacceptable tube fretting-wear at the supports. The fretting wear damage could ultimately cause tube failure. Such failures may require shut downs resulting in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting wear damage to tubes due to fluid excitations. Tubes in loose flat-bar supports have complex dynamics due to the possible combinations of geometry. The understanding of tube dynamics in the presence of this type of support and the associated fretting wear is still incomplete. These issues are addressed in this paper through simulations of the dynamics of tubes subjected to crossflow turbulence and fluidelastic instability forces. The finite element method is utilized to model the vibrations and impact dynamics. The tube model simulates a U-tube supported by 16 flat bars with clearances and axial offset. Results are presented and comparisons are made for the parameters influencing the fretting-wear damage such as contact ratio, impact forces and normal work rate. The effect of support clearance and support axial offset are investigated.

Fluidelastic Instability Modeling of Loosely Supported Multispan U-Tubes in Nuclear Steam Generators

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Tube Bundle ◽  
Design Guidelines ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Steam Generators ◽  
Fluidelastic Instability ◽  
Wear Damage

Steam generators in nuclear power plants have experienced tube failures caused by flow-induced vibrations. Two excitation mechanisms are responsible for such failures; random turbulence excitation and fluidelastic instability. The random turbulence excitation mechanism results in long-term failures due to fretting-wear damage at the tube supports, while fluidelastic instability results in short-term failures due to excessive vibration of the tubes. Such failures may require shutdowns, which result in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting-wear damage to tubes due to fluid excitation. In this paper, a numerical model is developed to predict the nonlinear dynamic response of a steam generator with multispan U-tubes and anti-vibration bar supports, and the associated fretting wear due to fluid excitation. Both the crossflow turbulence and fluidelastic instability forces are considered in this model. The finite element method is utilized to model the vibrations and impact dynamics. The tube bundle geometry is similar to the geometry used in CANDU steam generators. Eight sets of flat-bar supports are considered. Moreover, the effect of clearances between the tubes and their supports, and axial offset between the supports are investigated. The results are presented and comparisons are made for the parameters influencing the fretting-wear damage, such as contact ratio, impact forces, and normal work rate. It is clear that tubes in loose flat-bar supports have complex dynamics due to a combination of geometry, tube-to-support clearance, offset, and misalignment. However, the results of the numerical simulation along with the developed model provide new insight into the flow-induced vibration mechanism and fretting wear of multispan U-tubes that can be incorporated into future design guidelines for steam generators and large heat exchangers.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

A comparative study on the fretting wear of steam generator tubes in korean power plants

Wear ◽  
2003 ◽  
Vol 255 (7-12) ◽  
pp. 1198-1208 ◽  
Author(s):  
Young-Ho Lee ◽  
Hyung-Kyu Kim ◽  
Hong-Deok Kim ◽  
Chi-Yong Park ◽  
In-Sup Kim
Keyword(s):  
Comparative Study ◽  
Steam Generator ◽  
Power Plants ◽  
Fretting Wear ◽  
Steam Generator Tubes

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

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