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RISS 인기검색어
- 검색키워드
- 저자 : 박남규(Num-Kyu Park) (검색결과 2 건)
- 무료
- 기관 내 무료
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김해란(Hai-Lan Jin),이영신(Young-Shin Lee),이현승(Hyun-Seung Lee),박남규(Num-Kyu Park) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11
Each cell of spacer grids support the fuel rods continuously to maintain the spacing between the fuel rods in a nuclear reactor core and minimize the lateral displacement of fuel rods during operation. Although numbers of spacer grids support a fuel rod, the vibration behavior still exist in reactor due to coolant flow. So it is important to understand the fuel rod vibration characteristics in details. Because the spacer grid numbers greatly effect the fuel rod vibration characteristics, it is also important to know the fuel rod vibration characteristics with different number of spacer grid support conditions. In this pacer, modal simulations on fuel rod were done using the workbench 12.0 program to get nature frequencies and mode shapes under different number of spacer grid conditions. In simulation, real form fuel rod and spacer grids were used and assume the fuel rod contact tightly with the spacer grids. The mesh was performed by ICEM program and verified quality of mesh with mesh quality histogram that high quality of mesh can improve the simulation results. As a simulation results, at two supports(bottom and top spacer grid) conditions nature frequency got the most small value. At three supports(bottom, top and mid spacer grid) and four supports(bottom, top and two mid) conditions, nature frequencies were similar which 3 times higher than two support condition. At five supports(bottom, top and three mid) and seven(bottom, top, three mid and two IFM) conditions, the nature frequencies were almost same and 3 times higher than three and four supports conditions.
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냉각수 유동을 고려한 핵 연료봉의 유체-고체 연계 해석
김해란(Hai-Lan Jin),이영신(Young-Shin Lee),이현승(Hyun-Seung Lee),김영지(Young-Ji Kim),박남규(Num-Kyu Park),전경락(Kyung-Rok Jeon) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.3
This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Both fluid analysis of the coolant which flow in the axial direction and a structural analysis using the pressure data resulting from fluid analysis were carried out under different output flow velocity conditions. In the flow analysis, streamlines and maximum flow velocity at different domain were easily acquired from the analysis results. In the structure analysis, maximum stress of the nuclear reactor fuel rod was calculated and the location that has the maximum stress was found.
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