• Proceedings of the KSME Conference
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• 2004.11a
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• pp.256-261
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• 2004

To evaluate the effect of spring shape on the fretting wear of nuclear fuel rod, sliding wear tests were performed using three kinds of space grid springs in room temperature air and water. With increasing slip amplitude, wear volume of each spring gradually increased. It is apparently shown that spring with convex shape had a relatively high wear resistance compared with concave shape springs. It is suggested that the ratio of the wear volume to the worn area can be suggested as an efficient and valid parameter to evaluate the wear resistibility of a fuel grid spring.

• Journal of the Korean Society of Safety
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• v.19 no.2
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• pp.11-15
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• 2004

The fretting-fatigue by leaking is one of the significant degradation in steam generator tubes. In this study, the burst pressure of inclined damaged steam generator tubes were obtained from three criterions by using the finite element method. The analysis results were also compared with the experiment data from published references and they showed a good agreement with the experiment data.

• Tribology and Lubricants
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• v.22 no.1
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• pp.8-13
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• 2006

Sliding wear tests have been performed to evaluate the effect of normal load decrease on the wear depth of nuclear fuel rods in room temperature air. The objectives of this study are to quantitatively evaluate the supporting ability of spacer grid springs, to estimate the wear depth by using the contacting force decrease and to compare the wear behavior with increasing test cycles (up to $10^7$) at each spring condition. The result showed that the contacting load decrease depends on the spring shape and the applied slip amplitude. The estimated wear depth is smaller when compared with measured wear depth. Based on the test results, the wear mechanism, the role of wear debris layer and the spring shape effect were discussed.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.205-210
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• 2001

Fatigue crack growth tests were carried out using high manganese cast steel under constant amplitude loading. Average crystal grain sizes of the material are $200{\mu}m$ and $1000{\mu}m$. For this material, ${\Delta}K_{th}$ is about $8MPa{\sqrt{m}}$ which is quiet large as compared to the general structural steels and the crack growth rate is lower than the general structural steels especilly in the low ${\Delta}K$ regsion. The reason of this behavior is crack closure due to fracture surface roughness and fretting oxide. The relationship between da/dN and the ${\Delta}K_{eq}$ was represented by narrow band regardless of the stress ratio.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.11a
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• pp.59-62
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• 2001

Flow induced vibration in steam generators has caused dynamic interactions between tubes and contacting materials resulting in fretting wear . Series of experiments have been performed to examine the wear properties of Inconel 690 steam generator tubes in various environmental conditions. For the present study, the test rig was designed to examine the fretting wear and rolling wear properties in high temperature(room temperature - 290。C) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of the steam generator tube materials. To investigate the wear mechanism of material, the worn was observed using scanning electron microscopy. The weight loss increase at higher test temperature was caused by the decrease of water viscosity and the mechanical property change of tube material. The mechanical property changes of steam generator tube material, such as decrease of hardness or yield stress in the high temperature tests. From the SEM observation of worn surfaces, the severe wear scars were observed in specimens tested at the higher temperature.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.737-746
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• 2007

A spacer grid, which is one of the most important structural components in a PWR fuel assembly, supports its fuel rods laterally and vertically. Based on in-house design experience, scrutiny of the design features of advanced nuclear fuels and the patents of other spacer grids, KAERI has devised its own spacer grid shapes and acquired patents. In this study, a performance evaluation of KAERI's spacer grid shapes was carried out from mechanical/structural and thermohydraulic view points. A comparative performance evaluation of commercial spacer grid shapes was also carried out. The comparisons addressed the spring characteristics, fuel rod vibration characteristics, fretting wear resistance, impact strength characteristics, CHF enhancement, and the pressure drop level of the spacer grid shapes. The results show that the performances of KAERI's spacer grid shapes are as good as or better than those of the commercial spacer grid shapes.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.496-501
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• 1998

최근 국내의 PWR 발전소에서는 유체유발진동에 의한 핵연료의 Fretting Wear가 많이 발생하였다. 이는 Baffle Jetting이나 그 밖의 요인도 있을 수 있으나 핵연료의 장주기화, 높은 열적여유도등의 설계요건을 만족하기 위한 노심 내의 유동조건 변화에 기인한다. 특히 고리 2호기에서 발생한 핵연료 손상 중 15%정도가 유체유발진동으로 추정되고 있다. 따라서 본 연구는 손상 핵연료의 노심내 위치, 부위, 유동조건 등으로 부터 유체유발진동의 주요 손상 원인을 규명하는데 있다. 이를 위해 핵연료 집합체에서 발생할 수 있는 유체유발진동 메카니즘의 특징과 유동조건, 손상 핵연료의 노심내 위치, 파손 부위, 집합체와 지지격자의 기하학적 형태를 고려한 유동 방향 등을 연관 분석 결과 파손을 일으키는 주요원인을 단일 집합체 내에서 발생되는 Vortex Shedding과 인접한 집합체 사이에서 발생되는 Fluidelastic Instability의 중복효과로 규명하였다 또한 최근 핵연료 설계에 도입된 Mixing Vane의 효과가 과도하여 핵연료 손상을 일으키는 가설을 정립하였다.

• Proceedings of the KSME Conference
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• 2004.04a
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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.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.950-955
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• 2004

Slip displacement is brought into focus to study the tube fretting wear that occurs on the contact by the springs. An oscillating tube was in contact with plate support springs. The contact condition was varied as normal force 5 N, and gaps of 0.1 and 0.2 mm in the experiment. The oscillation range of the tube was also varied as 0.2, 0.3, 0.4 and 0.7 mm. Formulas for predicting the slip displacement range were derived in terms of the vibration amplitudes measured during the tube oscillation. It was found that the slip displacement in transverse direction was much higher ($720{\sim}33000$ times) than that in axial one. This resulted in the severer wear on the contact suffered from transverse slip.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2000.11a
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• pp.37-43
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• 2000

The effect of the alignment of the contacting bodies is concerned. Tilting of a body (indenter) is regarded as a mis-alignment. Recently developed method fur evaluating the contact normal traction is introduced and discussed, in which piecewise parabolic profile is considered as a generalized contact profile. Indentation by a wedge with rounded apex is solved as an example problem. In the case of partial slip regime, the variation of shear traction is examined during the closed path of shear force by using the efficient numerical technique. Concentration of the slip region is also investigated. By evaluating the energy dissipation from the contact surface, the effect of the mis-alignment-alignment on wear is discussed.