• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.12
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• pp.1653-1661
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• 2012

In this study, the pressure drop changes and structural characteristics of a SMART rod bundle under the effect of a coolant were investigated. The turbulence model of the BSL Reynolds stress model was used to model the coolant flow, and a fluid solid interaction simulation was conducted. First, fuel rod vibration analysis was performed to confirm the natural frequency of the fuel rod, which was supported by spacer grid assemblies, and this was compared with experimental results. From the experimental results, the natural frequency was found to be 48 Hz, and the error compared with the simulation results was 2%. The pressure drop at the rod bundle was calculated and compared with the experimental data; it showed an error of 8%, demonstrating the simulation accuracy. In the flow analysis, the flow velocity and secondary flow at different domains were calculated, and vortex generation was also observed. Finally, through the fluid solid interaction analysis, the fuel rod displacements caused by flow-induced vibrations were calculated. Then, calculated displacement PSD at maximum displacement happed point.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.260-260
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• 2003

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.847-848
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.285-285
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.169.2-169
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.236.1-236
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.194.2-194
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.182-182
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.304.2-304.2
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• 2003