• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.183.2-183
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.150.2-150.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.135.2-135
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• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.10a
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• pp.115.2-115.2
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• 2003

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.137-163
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• 1991

• 전기의세계
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• v.15 no.2
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• pp.8-14
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• 1966

보편적으로 들수 있는 상용에너지 공급원으로는 신탄, 석탄, 석유, 천연가스, 수력을 들 수 있고 최근에는 핵분열(Nuclear Fission)을 이용한 원자력이 장래 에너지 공급원으로 가장 유망시되고 있다. 이 밖에 아직 상용으로 각광을 받지 못하고 있지만 금후 개발잠재력(Development Potential)을 갖고 있는 조력, 풍력, 태양열, 지열등이 있다. 이중 풍력은 주로 농사용 양수동력에 조력, 태양열, 지열은 발전용으로 상당히 이용되고 있는 지감도 있다. 본 고에서는 일차에너지인 수력, 석탄, 석유 천연가스 핵자원등에 대하여 살펴보기로 한다.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.19-31
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• 1994

The objective of this study is to conduct a thermal-hydraulic analysis on the spent fuel pool and to evaluate a parametric effect for the thermal-hydraulic analysis of spent fuel pool. The selected parameters are the Reynolds Number and the gap flow through the oater gap between fuel cell and fuel bundle. The simplified flow network for a path of fuel cells is used to analyze the natural circulation phenomenon. In the flow network analysis, the pressure drop for each assembly from the entrance of the fuel rack to the exit of the fuel assembly is balanced by the driving head due to the density difference between the pool fluid and the average fluid in each spent fuel assembly. The governing equations ore developed using this relation. But, since the parameters(flow rate, pressure loss coefficient, decay heat, density)are coupled each other, iteration method is used to obtain the solution. For the analysis of the YGN 3&4 spent fuel rack, 12 channels are considered and the inputs such as decay heat and pressure loss coefficient are determined conservatively. The results show the thermal-hydraulic characteristics(void fraction, density, boiling height)of the YGN 3&4 spent fuel rack. There occurs small amount of boiling in the cells. Fuel cladding temperature is lower than 343.3$^{\circ}C$. The evaluation of parametric effect indicates that flow resistances by geometric effect are very sensitive to Reynolds number in the transition region and the gap flow is negligible because of the larger flow resistance in the gap flow path than in the fuel bundle.

• Nuclear industry
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• no.6
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• pp.40-50
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• 1979

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.276-290
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• 1988

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.47-50
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• 2002

The numerical simulation of a 271-rod fuel assembly of nuclear Liquid-Metal Fast Reactor (LMFR) with an infernal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it potentially has very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within a assembly with practically no change in mixed mean temperature at the assembly exit.