• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.67-72
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• 2004

Fundamentally, slip displacement should be known to solve the problem related to the fretting wear. For this, methods for measuring the slip displacement range in the contact surface of the tube and the supports were introduced and analyzed in this study. Also the increment of the slip displacement during a cycle was calculated using the revised formulas. As a result, the slip displacement newly evaluated was much higher ($7{\sim}50$ times) than that previously evaluated especially in the case of the gap existence. This enables to explain the severe wear found when there was a gap between the tube and the supports.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 1999년도 제30회 추계학술대회
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• pp.78-83
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• 1999

Fretting Wear Behavior and Characteristics were investigated with Inconel 690 tube which is used in Nuclear Power Plant. Fretting Wear Tester has been engaged in the study of characteristics on tube fretted induced to various factors, i.e. vibration amplitude, contact type, material properties load support geometry, environmental effect Test has been performed in air and at room temperature. Wear volume rate has been approximately estimated as a function of normal load.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2008년도 추계학술발표대회 및 제15회 신소재 심포지엄
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• pp.23.2-23.2
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• 2008

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.220-220
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• 1998

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.616-621
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• 2003

Fluid-elastic instability is believed to be a cause of the large-amplitude vibration and resulting rapid wear of heat exchanger tubes when the flow velocity exceeds a critical value. For sub-critical flow velocities, the random turbulence excitation is the main mechanism to be considered in predicting the long-term wear of steam generator tubes. Since flow-induced interactions of the tubes with tube supports in the sub-critical flow velocity can cause a localized tube wear, tube movement in the clearance between the tube and tube support as well as the normal contact force on the tubes by fluid should be maintained as low as possible. A simplified method is used for predicting fretting-wear damage of the double $90^{\circ}$U-bend tubes. The approach employed is based on the straight single-span tube analytical model proposed by Connors, the linear structural dynamic theory of Appendix N-1300 to ASME Section III and the Archard's equation for adhesive wear. Results from the presented method show a similar trend compared with the field data. This method can be utilized to predict the fretting-wear of the double $90^{\circ}$U-bend tubes in steam generators.

• Tribology and Lubricants
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• 제24권6호
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• pp.362-367
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• 2008

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. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• 대한기계학회논문집A
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• 제34권1호
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• pp.49-54
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• 2010

탄성변형 접촉 조건을 가지는 이중냉각 핵연료봉과 이를 지지하는 지지구조체 사이에서 발생하는 프레팅 마멸거동을 모사 시험편을 이용하여 실험적으로 분석하였다. 이중냉각 핵연료봉은 기존의 핵연료에 비해 외경이 증가하므로 새로운 형상을 가지는 지지구조체의 적용이 필수적이며 이에 대한 진동 특성 및 마멸 저항성에 대한 평가가 필수적이다. 본 연구에서는 현재까지 제안된 다양한 형상 중에서 대표적으로 엠보싱 형상을 가지는 지지구조체의 모사 시험편을 이용하여 이중냉각 핵연료봉의 내마멸 특성을 분석하였다. 개발된 지지구조체 특성 시험장비를 이용하여 모사된 지지구조체 시험편의 특성시험을 수행하였으며 이를 해석에 의한 결과와 비교하였다. 또한 기존의 핵연료 내마멸시험과 동일한 조건 및 장비를 이용하여 프레팅 마멸시험을 수행하여 이중냉각 핵연료봉의 프레팅 마멸거동을 관찰하였다. 본 논문에서는 실험결과로부터 지지구조체 특성과 프레팅 마멸거동 사이의 상관관계에 대하여 자세히 논하였다.

• 열처리공학회지
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• 제19권1호
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• pp.30-36
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• 2006

Fretting fatigue behavior of SCM420 steel commonly used in the automotive industry for structural applications was investigated in this study. In addition, the effect of bridge pad on the fretting fatigue test was evaluated from different pad materials and following conclusions were drawn. Simple fatigue limit of SCM 420 steel was determined to be 350 MPa while this value was 225 MPa and 285.5 MPa with SCM420H and with SM45C pad, respectively. Reduction in fatigue limit was, thus found to be 35.7% and 17.9% with SCM 420H pad and SM45C pad, respectively. Results of fracture surface observation revealed that typical striation pattern of fatigue failure existed as well as dimpled and cleavage frature appearance was found in final fractured region. From the EDS compositional analysis, test sample and pad part all had high signals for oxygen and iron, suggesting that worn particles might be iron oxide, although exact chemical composition has to be confirmed. Considerable reduction in fatigue life was apparent in SCM 420 steel under fretting fatigue against simple fatigue. Such reduced fatigue life by fretting damage should be considered as an important factor not only in the viewpoint of repairing but also inevitably in the design stage of structural components.

• KSTLE International Journal
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• 제2권2호
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• pp.85-92
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• 2001

The variation of contact traction induced by different contact shapes is studied experimentally and theoretically. Considerations fer the contact shape are rounded, truncated and truncated with rounding punches. A fretting wear experiment is conducted with the contact configuration of the strip on the tube specimens. The strip specimen is pressed to form the end profile of a rounded and truncated with rounding punches shape. Wear on the tube is investigated, which is regarded as the slip region of the contact surface. Taken into consideration is the general solution of the normal traction in the case of the indentation by a punch with its end profile of the combination of parabolas. Then, partial slip solution is obtained numerically, which is compared with the wear on the tube. The radius of the rounding and the obliquity of the edge truncation affect the tractions considerably. It is found that the proper choice of the end profile can restrain the contact-induced failure such aswear.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2007년도 정기 학술대회 논문집
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• pp.145-150
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• 2007

The spacer grid set is a component in the nuclear fuel assembly. The set supports the fuel rods saftely. Therefore, the spacer gl1d set should have sufficient strength for the external impact forces. The fretting wear occurs between the spring of the fuel rod and the spacer grid due to tile flow-induced vibration. The conceptual design of the spacer grid set is performed based on the Independence Axiom of axiomatic design. Two functional requirements are defined and corresponding design parameters are selected. The overall flow of the design is defined according to the application of axiomatic design. The design for the impact load is carried out by using nonlinear dynamic analysis to determine the length of the dimple. Topology optimization is carried out to determine a new configuration of the spring. The fretting wear is reduced by shape optimization using the homology theory. In the design to reduce the fretting wear, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the shape optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow a slight plastic deformation. Shape optimization results are confirmed through nonlinear static analysis because the contact area becomes wider.