DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR

2010 ◽  
Vol 24 (15n16) ◽  
pp. 2603-2608
Author(s):  
CHOON YEOL LEE ◽  
JOONG HO KIM ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Nuclear Power ◽  
Contact Region ◽  
Water Environment ◽  
Control Unit ◽  
Fretting Wear ◽  
Test Rig ◽  
Realistic Condition ◽  
Wear Simulator

In nuclear power plant, fretting wear due to a combination of impact and sliding motions of the U-tubes against the supports and/or foreign objects caused by flow induced vibration, can make a serious problem in steam generator. A test rig, fretting wear simulator, is developed to elucidate fretting wear mechanism qualitatively and quantitatively. The realistic condition of steam generator of high temperature up to 320°C, high pressure up to 15 MPa, and water environment could be achieved by a test rig. The fretting wear simulator consists of main frame, water loop system, and control unit. Actual contact region under a realistic condition of steam generator was isolated using autoclave. Effects of various parameters such as the amounts of impact and sliding motions, applied loads and initial gaps and so forth are considered in this research. After the experiment, wear damage was measured by a three-dimensional profiler and the surface was also studied by SEM microscopically. Initial results were also presented.

ERRATUM: "DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR"

2011 ◽  
Vol 25 (20) ◽  
pp. 2789-2789
Author(s):  
CHOON YEOL LEE ◽  
JOONG HO KIM ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Fretting Wear ◽  
Wear Simulator

Experimental Investigation of the Flow-Induced Vibrations of a Rod Cluster Control Assembly Inside Guides With Enlarged Gaps

2019 ◽  
Author(s):  
Pierre Moussou ◽  
Vincent Fichet ◽  
Luc Pastur ◽  
Constance Duhamel ◽  
Yannick Tampango
Keyword(s):  
Experimental Investigation ◽  
Flow Velocity ◽  
Nuclear Power ◽  
Axial Flow ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Test Rig ◽  
Pressurized Water ◽  
Control Assembly ◽  
Control Rods

Abstract In order to better understand the mechanisms of fretting wear damage of guide cards in some Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP), an experimental investigation is undertaken at the Magaly facility in Le Creusot. The test rig consists of a complete Rod Cluster with eleven Guide Cards, submitted to axial flow inside a water tunnel. In order to mimic the effect of fretting wear, the four lower guide cards have enlarged gaps, so that the Control Rods are free to oscillate. The test rig is operated at ambient temperature and pressure, and Plexiglas walls can be arranged along its upper part, and a series of camera records the vibrations of the control rods above and below the guide cards. The vertical flow velocity is in the range of a few m/s. Beam-like pinned-pinned modes at about 5 Hz are observed, and oscillations of several mm of the central rods are measured, which come along with impacts at the higher flow velocities. A simple non-linear calculation reveals that the main effect of the impacts between Control Rods and Guide Cards is an increase of the natural frequency of the rods by about 10%. Furthermore, as the vibration spectra collapse remarkably well with the flow velocity, the experiments prove that turbulent forcing is responsible for the large oscillations of the control rods, no other mechanism being involved.

The Fretting Wear Characteristics of CrN and TiN Coating on Steam Generator Tube

2010 ◽  
Author(s):  
Jin-Seon Kim ◽  
Joo Hoon Choi ◽  
Young-Ze Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Relative Displacement ◽  
Fretting Wear ◽  
Improve Performance ◽  
Tin Coating ◽  
Stick Slip ◽  
Steam Generator Tube ◽  
Lower Load ◽  
Flow Induced Vibrations

A steam generator tube of a nuclear power plant is damaged by a fretting phenomenon caused by flow induced vibrations (FIV). In this work, the surface of the tube was coated with CrN or TiN as a measure to improve performance of the fretting wear resistance. Fretting wear regime was classified by determining a phase difference between friction and relative displacement signals and contact characteristics were analyzed. As a result, coating increased the friction coefficient. At a lower load, contact condition shifted from gross slip to stick slip.

309 High Temperature Impact-Fretting Wear Experiments for Evaluation of Pipe Wall Thinning Phenomena in Nuclear Power Plants

2013 ◽  
Vol 2013.49 (0) ◽  
pp. 87-88
Author(s):  
Yoshiki SATO ◽  
Akira IWABUCHI ◽  
Michimasa UCHIDATE ◽  
Hitoshi YASHIRO ◽  
Akito OYAKAMA ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pipe Wall ◽  
Fretting Wear ◽  
Wall Thinning ◽  
Temperature Impact

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.

Experimental Damping of a Heat Exchange Tube With a Large Number of Supports in Air and Water

2017 ◽  
Author(s):  
Tae-Jung Park ◽  
Chang-Hoon Ha ◽  
Min-Ki Cho ◽  
Heung Seok Kang ◽  
Kang Hee Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Large Scale ◽  
Damping Ratio ◽  
Water Environment ◽  
Random Force ◽  
Flow Induced Vibration ◽  
Heat Exchange Tube ◽  
Half Power ◽  
Support Grid

The flow induced vibration occurs frequently in a steam generator in the nuclear power plant. The large-scale steam generator has a large number of tube supports whose cell has rhombus-type shape, and there is a tiny clearance between tube and its support grid. The damping is very complex because of non-linearity and randomness. The experiment for damping was performed to investigate it with a number of 13 support spans both in air and water environment. The lower part of multi-span fixture was excited by root-mean-square random force with the range of 1∼10 newton to get the frequency response function. The half-power bandwidth method was applied to obtain the damping ratio. The sensitivity of a number of spans was investigated in the range of 9 ∼ 13. In addition, the damping was reviewed from a comparison with Pettigrew [1∼4] and ASME B&PV Code [5].

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.

Fretting Wear Characteristics of INCONEL 690 and INCONEL 600 in Water Environment

2005 ◽  
Vol 297-300 ◽  
pp. 1412-1417 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Chi Yong Park ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Water Environment ◽  
Fretting Wear ◽  
Wear Loss ◽  
Stick Slip ◽  
Inconel 600 ◽  
Delamination Wear ◽  
Slip Regime ◽  
Inconel 690 ◽  
Steam Generator Tubes

Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are INCONEL 690 (I-690) and INCONEL 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

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