• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.110.1-110
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.347.1-347
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.206-206
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.81-81
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• 2002

• Proceedings of the KIEE Conference
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• 1983.07a
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• pp.120-122
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• 1983

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.395-396
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• 2017

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.4 s.70
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• pp.405-416
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• 2005

This study contains the static buckling tests and static buckling analyses for small size grids and full size grids. The buckling tests and finite element analyses were performed to evaluate the buckling characteristics of the spacer grids in a pressurized water reactor fuel assembly and to evaluate the possibility of the prediction lot the buckling strength of spacer grids. The buckling tests were performed for small size grids and full size grids, and the correlations between buckling strength and the number of straps and the correlations between buckling strength and the number of rows are derived based on the test results. The static buckling analyses were performed to identify the effect of the number of rows and the number of columns on the buckling strength of spacer grid by a finite element method using ANSYS program and the results were compared with the buckling test results.

• Journal of the Korean Society of Industry Convergence
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• v.3 no.3
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• pp.201-207
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• 2000

In this study, methodologies for checking the interference between in-core instrumentation (ICI) guide tubes for routing of ICI guide tubes in the reactor coolant system of typical Pressurized Water Reactor under cold and normal operation (NOP) conditions are presented. The closest points of ICI guide tubes under cold condition are calculated by using minimize technique and are used as data for NOP analysis. Movements of ICI guide tubes under NOP condition are performed by the commercial computer code, SUPERPIPE.

• Structural Engineering and Mechanics
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• v.5 no.2
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• pp.149-165
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• 1997

A dynamic time-history analysis of the coupled internals and core in the vertical direction is performed as a part of the fuel assembly qualification program. To reflect the interaction between the fuel rods and grid cage, friction element is developed and is implemented. Also derived here is a method to calculate a hydraulic force on the reactor internals due to pipe break. Peak responses are obtained for the excitations induced from earthquake and pipe break. The dynamic responses such as fuel assembly axial forces and lift-off characteristics are investigated.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.