• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.424-429
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• 1993

Numeric characters are printed at the end part of nuclear fuel rod containing nuclear pellets. Fuel rods are discriminated and managed systematically by these characters in the process of producing fuel assembly. The characters are also used to examine manufacturing process of fuel rods in the survey of burnup efficiency as well as in inspection of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies in automatic manufacture of fuel assembly. In this study, character recognition system is developed. In the developed system, mesh feature extracted from each character written in the fuel rod has been compared with reference feature value stored in database, and the character is thus identified. In the result of experiment, 95.83 percent recognition rate is achievable.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.760-763
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• 2001

In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembling processes. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the processes involved in the spent fuel handling and disassembling precesses are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator to enhance the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimization of the main processes and maintenance processes of the spent fuel management.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.349-350
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• 2010

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.89.1-89
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.10a
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• pp.187-188
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• 2016

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.339-343
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• 1995

PWR 사용후핵연료 집합체를 운반할 수 있는 수송용기를 개발하기 위하여 단면이 수송용기의 실제 크기인 slice 모델을 사용하여 법규에서 규정하고 있는 정상조건인 주변온도 38$^{\circ}C$에서 냉각 매체로 nitrogen 과 helium 인 경우에 대하여 열시험을 수행하여 수송용기의 열전달 특성 및 핵연료봉의 건전성을 평가하였다. 열시험결과 내부핵연봉의 최대 은도는 각각 448$^{\circ}C$ 와 416$^{\circ}C$로 측정되었다. 이 값들은 핵연료봉의 건전성 유지에 필요한 허용치 이내 만족하는 것으로 수송 용기의 열전달성능이 우수함을 입증하는 것이다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.83-84
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• 2012

SF 집합체 해체장치 개념설계요건 설정을 위해서 PWR $16{\times}16$ SF집합체 제원을 분석하였다. 또한 집합체 해체장치 주요요건을 도출하여 핵심메커니즘을 도출하였다. 주요요건은 다음과 같다. 집합체의 최대 clamping 힘은 각 grid의 경우: 240 kg, 하부노즐의 경우: 900 kg이다. 3 축 방향에서 절단을 위한 정확한 위치공자는 ${\pm}0.25mm$이다. 또한 처분을 위해 cuttings, fines 및 다양한 hardware를 수거하는 기능을 제공해야 한다. SF 집합체 해체를 위하여 드릴링 방식을 채택하였다. PWR SF의 종류에 따라 드릴링 위치가 다르기 때문에 위치제어와 해제장치 하단과 중간에 있는 X, Y, Z 제어를 할 수 있는 구조로 고안 하였다. SF 집합체 해체장치는 국내에서 가동되는 모든 PWR SF 집합체를 해체할 수 있는 구조로서 범용성을 가지고 있다. 원격 유지보수성을 향상하기 위하여 Solid Works 프로그램 툴(tool)을 이용하여 8개의 주요 모듈을 구성하였고, SF 집합체 해제장치 개념을 3D로 설계하였다.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.352.1-352
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.261-267
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• 2004

Inventories, projections, and characteristics of spent nuclear fuel(SNF) generated from domestic nuclear power plants were updated to support high-level waste disposal system design. The historical and projected inventory by the end 2055 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively The ratio of quantity for TEX>$17{\times}17$ SNF was shown to be 0.6 as of 2003. The amount of TEX>$17{\times}17$ SNF, however, will be less than that of TEX>$16{\times}16$ KSFA after 2012, while the quantity of TEX>$16{\times}16$ KSFA will reach to 70% of the total spent fuels in the 2055. Average turnup of SNF revealed ~36GWD/MTU and ~40GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will exceed 45GWD/MTU at the end of 2000's. Therefore, it seems reasonable to use the TEX>$17{\times}17$ 4.5w/o, 45GWD/MTU as the Reference SNF at present state. The TEX>$16{\times}16$ KSFA 4.5w/o, 55GWD/MTU, however, should be Reference SNF after ~2010.