• Title/Summary/Keyword: 모의핵연료

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Specific Heat Characteristics of Ceramic Fuels (산화물핵연료의 비열특성)

  • Kang Kweon Ho;Park Chang Je;Ryu Ho Jin;Song Kee Chan;Yang Myung Seung;Moon Heung Soo;Lee Young Woo;Na Sang Ho
    • Journal of Energy Engineering
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    • v.13 no.4
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    • pp.259-266
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    • 2004

  • Specific heat mechanism of oxide fuel is contributed by lattice vibration, dilatation, conduction electron and defect and excess specific heat. Model of oxide fuel for specific heat consists of specific heat at constant pressure term, dilatation specific heat term and defect specific heat term. In this study experimental and published data on the specific heats of oxide nuclear fuels have been reviewed and analyzed to recommend the best fitting model. The oxide fuels considered in this paper were UO$_2$, mixed (U, Pu) oxides and spent fuel. The specific heat data of spent fuel has been replaced by that of simulated fuel.

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Dimensional Measurement of Spent Fuel Assemblies Using Image Processing Technique (영상처리기술에 의한 사용후핵연료 집합체의 제원 측정)

  • Koo, Dae-Seo;Park, Seong-Won
    • Journal of the Korean Society for Nondestructive Testing
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    • v.22 no.1
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    • pp.9-13
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    • 2002

  • A pool image processing measurement method has been developed to improve the examination efficiency and to minimize the errors of dimensional measurements of spent fuel assemblies in pool. Diameter and length measurements of mock-up fuel rods using the image processing system are $-0.24{\pm}0.03mm,\;0.34{\pm}0.06mm$ on the basis of the true value and their maximum errors are within -0.3 and 0.4mm, respectively, According to the result of dimensional measurement of spent fuels in pool, the upper and lower part diameter and mid part diameter of fuel rods of the J44 fuel assembly irradiated for 2 cycles in the Kori-2 nuclear reactor were decreased by about 2.0 and 3.0% in comparison with design values, respectively. The length of fuel rods was elongated by about 0.4%. The change behavior of diameter and length. of fuel rods of the F02 fuel assembly irradiated for 3 cycles in the Kori-1 nuclear reactor showed a trend similar to the results of J44.

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MELCOR코드를 이용한 PHEBUS FPT-1 실험해석

  • 조성원;홍성완
    • Proceedings of the Korean Nuclear Society Conference
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    • 1997.10a
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    • pp.726-731
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    • 1997

  • 중대사고시 핵연료와 핵분열생성물의 거동을 파악하기 위한 PHEBUS FPT-1실험을 MELCOR 코드로 해석함으로써 코드의 모의 능력 및 실험의 최근 연구 동향과 측정의 타당성을 파악할 수 있었다. 노심을 포함한 전 계통의 열수력 거동에 대한 모의 결과는 측정 자료와 비교ㆍ분석하여 매우 타당한 결과를 얻은 것으로 판단되었다.

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