• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2047-2052
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• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.84-89
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• 2004

When the tube contacted to support, anti-vibration bar of the steam generator in nuclear power plant, the contact area is worn out by their relative displacement and contact force. Connors and Au-Yang found the relation between tube worn displacement and volume, or normal work rate at given gap size. The present analysis is obtained the relation between tube worn displacement and normal work rate at various gap size modifying Au-Yang's result. The results are compared with Connors and Yettisir and Pettigrew's results. The comparison shows that Yettisir and Pettigrew result is fairly good agreement with Connors and present results with gap clearance, 0.015in.

• Tribology and Lubricants
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• 제23권3호
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• pp.130-133
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• 2007

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 and support. The fretting wear of tube and 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.

• Tribology and Lubricants
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• 제24권3호
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• pp.129-132
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• 2008

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 and support. The fretting wear of tube and 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.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 제35회 춘계학술대회
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• pp.63-70
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• 2002

Wear on the tube-to-spring contact is investigated experimentally. The wear is caused by the vibration of the tube while the springs support it. As for the supporting conditions, applied are the contacting normal force (P) of 5 N, just-contact (P = 0 N) and the gap of 0.1 mm. The gap condition is tried far considering the influence of simultaneous impacting and sliding on wear. Results show that the wear volume increases in the order of the gap, the just-contact and the 5 N conditions. This is explained from the contact geometry of the spring, which is convex of smooth contour. The contact shear force is regarded smaller in the case of the gap existence compared with the other conditions. Wear mechanism is considered from SEM observation of the worn surface. The variation of the normal contact traction is analysed using the finite element analysis to estimate the slip displacement range on the contact with consulting the fretting map previously obtained.

• 한국소음진동공학회논문집
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• 제14권4호
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• pp.327-335
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• 2004

본 연구에서는 가동중인 원자력발전소의 이물질에 의한 프레팅 마모 특성을 평가하였다. 다양한 조건의 나선형 튜브 고유진동수 및 모드 형상을 구하기 위하여 모드 해석을 수행하였다. 이물질에 의한 나선형 튜브의 마모율을 Archard 공식을 이용하여 계산하였고 이로부터 튜브의 잔여 수명을 예측하였으며 튜브의 진동 특성이 잔여 수명에 미치는 영향을 고찰하였다. 또한 튜브에 가해지는 외압이 진동 및 프레팅 마모 특성에 미치는 영향을 평가하였다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 1999년도 제29회 춘계학술대회
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• pp.310-316
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• 1999

The fretting wear characteristics of the contact between Zircaloy-4 tube and Inconel 600 tube have investigated. Zircaloy-4 is used for fuel rod in nuclear reactor and Inconel 600 is used for tube In steam generator of nuclear power plant. A fretting wear tester was designed to be suitable for this fretting test. In this study, the number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. This study shows that the wear scar length of Zircaloy-4 and Inconel 600 increases as number of cycles, normal load and slip amplitude increase and the wear scar length of Zircaloy-4 is more longer than that of Inconel 600 due to the surface hardness.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 제35회 춘계학술대회
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• pp.88-95
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• 2002

Fretting of fuel rod cladding material, Zircaloy-4 Tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to bean ideal solution to fretting damage since fretting is closely related to wear, corrosion and fatigue. Therefore, in this study the fretting wear experiment was peformed using TiAIN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaioy-4 tube as one of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAIN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and the fretting wear mechanisms were delamination and plastic flow following by brittle fracture at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher slip amplitude.

• Tribology and Lubricants
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• 제24권4호
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• pp.186-189
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• 2008

The flow induced vibration in the nuclear fuel assembly causes the fretting wear between the fuel cladding tubes and the supporting grids. The reduction in tube thickness due to the fretting wear could be related to the serious damage on nuclear fuel assembly. In this paper, the effect of the water flow on fretting wear of nuclear fuel cladding tube against supporting grid was investigated through the fretting wear tester with water spout equipment. The test results were compared with the data conducted in the stationary water. At stationary water environment the wear debris was trapped between fretting surfaces, and then the fretting wear occurred by three-body abrasion. However, in the case of water flow, the two-body abrasive wear was the dominant wear mechanism, because the wear debris was easily removed by water flow.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.178-188
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• 2003

The wear of steam generator tubes is due to the vibration occurred between tubes and tube supporters. To predict the future wear depth, the wear constants of the impact and the sliding model is used. The wear constants, 3C/2 and K/3H, are found inversely from known wear depth and time. Using these constants, the future wear depths are found from two bodies that deform the elliptical shape. The results are compared with the measured wear depth of steam generator tubes in a nuclear power plant. The results show that the predicted wear depth envelopes the measured wear depth.