• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.04a
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• pp.295-296
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• 2014

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2009.04a
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• pp.614-615
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• 2009

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.739-740
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• 2012

• Proceedings of the SAREK Conference
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• 2005.06a
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• pp.131-131
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.133-133
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• 1999

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.65-66
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• 2011

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.488-493
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• 1998

전도, 복사 및 대류 열전달을 모두 고려하여 VSC-24 핵연료 저장용기 각부위의 온도분포 및 Vent내의 유동을 해석할 수 있는 열전달해석 프로그램을 작성하였다. 기존의 저장용기 해석에서 여러 전산프로그램을 이용하여 온도분포를 구한 것과는 달리 핵연료집합체 내부의 복사 및 대류 열전달을 포함하였으며 MSB와 VCC를 분리하지 않고 PHOENICS 전산코드만을 이용하여 저장용기 전체에 대한 해석을 실시하였다 프로그램 검증은 저장용기에 대한 기존의 결과 및 법적규제치와 비교하였으며 만족할 만한 결과를 보여 주었다. 이를 토대로 몇가지 경우의 조건에 대한 열전달해석을 실시하였다.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.746-751
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• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized water reactor are reviewed from the mechanical/structural point of view. And mechanical/structural tests and numerical analyses on the various spacer grid candidates that has been uniquely designed by KAERI are carried out to find out their mechanical/structural performance. As a result, the results from the numerical analyses are good agreements with test results.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.