Best Estimate Evaluation of Steam Line Break Accident Using Uncertainty Quantification Method

• Lee C. S.;Jin Y. K.;Kim S. W.;Choi C. J.;Lee S. Y.;Seo J. T.
• • Proceedings of the Korean Nuclear Society Conference
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  • 2003.10a
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  • pp.131-131
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  • 2003
• PDF

Performance Analysis of Westinghouse 3-Loop Plants with CENTS Code

• Song, In-Ho;Sohn, Suk-Whun;Huh, Jae-Young;Lee, Gyu-Cheon;Sohn, Jong-Joo;Kim, Eun-Kee
• • Proceedings of the Korean Nuclear Society Conference
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  • 2005.10a
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  • pp.1083-1084
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  • 2005
• PDF

Non-Linear Structural Analysis of Reactor Cavity Permanent Pool Seal Assembly

• Kim, M.G.;Kim, T.H.;Lee, D.H.;Kim, I.Y.;Choi, T.S.
• • Proceedings of the Korean Nuclear Society Conference
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  • 2004.10a
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  • pp.1357-1358
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  • 2004
• PDF

Evaluation of Decay Heat Distribution in the OASIS-32D Cask

• Kim, Yong IL;Han, Young Tae;Ahn, Joon Gi;Song, In Ho;Lee, Gyu Cheon
• • Proceedings of the Korean Radioactive Waste Society Conference
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  • 2018.05a
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  • pp.33-34
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  • 2018
• PDF

Study on Minimization of Weld Deformation for Core Shroud by Application of Design Optimization Technique

• Kim, J. G.;Park, K. W.;J. S. Moon;Y. H. Cho;J. H. Bae
• • Proceedings of the Korean Nuclear Society Conference
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  • 1999.10a
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  • pp.355.1-355
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  • 1999
• PDF

Transient Analysis of Charging System with Centrifugal Charging Pumps

• Park, H.S.;J.S. Ahn;C.K. Chung;Kim, E.K.;T.S. Ro
• • Proceedings of the Korean Nuclear Society Conference
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  • 1998.10a
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  • pp.117-117
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  • 1998
• PDF

Transient Analysis of Letdown System with Letdown Orifices

• J. S. Ahn;Park, H. S.;C. K. Chung;Kim, E. K.;T. S. Ro
• • Proceedings of the Korean Nuclear Society Conference
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  • 1999.05a
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  • pp.93-93
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  • 1999
• PDF

Improvement of the Reactor Vessel Closure Head Insulation for the Visual Inspection in the Korean Standard Nuclear Power Plant

• Kim, Seoug-Beom;Chung, Chang-Kyu;Lim, Duck-Jae;Kim, Hyun-Min;Cho, Kwang-Hee;Seo, Jong-Tae
• • Proceedings of the Korean Nuclear Society Conference
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  • 2005.10a
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  • pp.963-964
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  • 2005
• PDF

Development of a Multi-dimensional two Phase Thermal Hydraulic Analysis Code, MEDUSA, and its application to Feed Line Break Analysis

• Park, Chan-Eok;Lee, Sang-Il;Lee, Sang-Yong;Kim, Shin-Whan;Seo, Jong-Tae
• • Proceedings of the Korean Nuclear Society Conference
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  • 2004.10a
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  • pp.427-428
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  • 2004
• PDF

A Study on the Improvement of RCS Leak Rate Calculation Program

• Choi, N.I.;Suh, S.K.;Cho, J.H.;Yune, S.J.;Baik, S.J.;Ko, Y.B.;Shin, D.J.
• • Proceedings of the Korean Nuclear Society Conference
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  • 2004.10a
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  • pp.663-664
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  • 2004
• PDF

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