• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.209.2-209
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• 2002

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제34회 추계학술대회 개최
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• pp.245-251
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• 2001

A fretting wear tester is developed for experimental study on the fuel fretting problem of light water reactor. The feature of the developed tester is it can simulate the existence of gap between spring and fuel rod as well as different contacting force including the just-contact condition (0 N on the contact). Used are a servo-motor, an eccentric cylinder and lever mechanism for driving system. A spacer grid cell is constituted with four strap segments (each segment has a spring). This fretting wear tester can also be used as a fatigue tester of a spacer grid spring with the frequency of more than 10 Hz. It is required to simulate the frequency of the vibrating fuel rod due to flow-induced vibration in a reactor. In fretting wear test, up to two span-length of a fuel cladding tube can be accommodated. A specimen of cladding tube of one span-length is specially designed, which can be extended for two-span test. For .fatigue test, a device for clamping the spring fixture is installed additionally, Presently, the tester is designed for the condition of air environment and room temperature. The variation of the reciprocal distance is measured to check the stability of input force, which will be exerted to the cladding (for fretting wear. test) and the spring (for fatigue test) specimen.

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.17-26
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• 1996

프레팅마모 기인 연료봉 손상을 방지할 수 있는 노내 연료봉 지지조건은 잔여 지지격자스프링 변위량 또는 연료봉 /지지격자 갭에 의해 평가될 수 있다. 핵연료 설계 인자들이 프레팅마모 손상에 미치는 영향을 평가하기 위해 연소도의 함수로서 노내 연료봉 지지조건을 모사할 수 있는 방법론을 사용하여 GRID-FORCE프로그램을 개발하였다. 이 프로그램에서는 노내 연료봉 지지조건에 영향을 주는 주요 인자로서 피복관 크립, 초기 스프링 변위, 초기 스프링힘 그리고 스프링힘 조사이완이 고려된다. 이 주요 인자들에 대한 민감도 분석 결과, 초기 스프링 변위, 스프링힘 조사이완, 피복관 크립 순으로 노내 연료봉 지지조건에 영향을 주는 것으로 나타났다. 이 프로그램을 실제 노내에서 발생한 프레팅마모 기인 연료봉 손상에 적용한 결과를 토대로 판단해 볼 때 이 프로그램을 새로 개발된 피복관 재질 및 /또는 새로 개발된 지지격자 설계가 프레팅마모 기인 연료봉 손상을 방지할 수 있는 설계여유도를 효과적으로 평가할 수 있음을 알 수 있다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.112.2-112.2
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.328.2-328
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.258.2-258.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.254.1-254
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• 1999

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.195-196
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• 2002

Studied is fretting wear behaviour of transversely vibrating tube which is supported by springs and dimples. This simulates the fuel rod fretting due to flow-induced vibration in a nuclear reactor. The contact between spacer grid springs and fuel cladding tubes arc brought into focus in this paper. From the mechanical viewpoint, a concave contact shape of spring is considered to perform a wider distribution of the contact stress. Sliding/impacting experiments are conducted in air at room temperature with the conditions of positive contact force and gap existence to accommodate the mechanical condition between the fuel rod and the grid spring during reactor operation. It is found that wear region is separated and wear volume becomes larger as the supporting condition becomes poorer. Spring and dimple cause similar wear.

• KSTLE International Journal
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• 제4권1호
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• pp.8-13
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• 2003

Abrasion in fretting wear mechanism is studied experimentally with the specimens of two different shapes of spacer grid spring and fuel tubes of a nuclear fuel. Reciprocating sliding wear test has been carried out in the environment of air and water at room temperature. Especially, third body abrasion is referred to for explaining the wear region expansion found during the slip displacement increase with constant normal contact farce. It is found that the expansion behaviour depends on the contact shape. The small clearance between the tube and spring seems to be the preferable region of the wear particle accumulation, which causes third body abrasion of the non-contact area. Even in water environment the third body abrasion occurs apparently. Since the abrasion on the clearance contributes wear volume, the influence of the contact shape on the severity of third body abrasion should be considered to improve the grid spring design in the point of restraining wear damage of a nuclear fuel.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.73-82
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• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.