• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.145-145
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.309.2-309.2
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• 2003

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.911-916
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• 2012

It is important to identify dynamic characteristics of nuclear fuel components. Since the fuel always exposed to turbulent flow, the dynamic contact between grids and rods is one of the fuel failure modes. The dynamic behavior of grids in nuclear fuels is quite complex, since two pairs of spring support are placed in the limited space. The strap in a cell has single spring and double dimples and this paper focuses on investigation of the grid strap(Test Fuel Strap, TFS) vibration in one cell. To identify the grid strap vibration, modal analysis of the strap is performed using Finite Element Method (FEM). Modal testing on a $5{\times}5$ grid structure without rods is performed. The modal testing results are compared to analytic results. In addition, random test considering rod effect is performed about a $5{\times}5$ grid with rods under real contact condition in the air. Finally, the strap vibration of a $5{\times}5$ fuel bundle in INvestigation of Flow INduced vIbraTion(INFINIT) facility is measured in real fluid velocity condition without heating. It is shown that modal frequencies from the test are almost equal to those peak frequencies in the INFINIT test.

• Journal of the Korean Society of Combustion
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• v.23 no.1
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• pp.44-54
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• 2018

Heat and mass balance process modeling has been conducted for a coal-firing power plant to be used as a testbed facility for development of various plant systems and equipment. As the material and design of the boiler tube bundle and fuel conversion to the biomass have become major concerns, the process modeling is required to incorporate those features in its calculation. The simulation cases for two different generation load show the satisfying results compared to the operational data from the actual system. Based on the established process conditions, the hypothetical case using wood pellet has also been simulated. Additional calculations for the tube bundle has been conducted regarding the changes in the tube material and design.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.815-824
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• 2016

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.109-114
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• 1995

A preliminary study of the heterogeneity effect of spent P% fuel in CANDU was made using a reduced spent PWR fuel data base. The instantaneous core simulation has shown that the refueling ripple in the CANDU reactor is large if the spent PWR fuel is directly used. But the fuel heterogeneity effect can be reduced appreciably by blending spent PWR fuel with a small amount of fresh UO$_2$. The refueling simulation has shown that the operating margins of 6.0% and 8.7% are achievable for the peak channel and bundle powers, respectively, with the blended fuel.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.19-31
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• 1994

The objective of this study is to conduct a thermal-hydraulic analysis on the spent fuel pool and to evaluate a parametric effect for the thermal-hydraulic analysis of spent fuel pool. The selected parameters are the Reynolds Number and the gap flow through the oater gap between fuel cell and fuel bundle. The simplified flow network for a path of fuel cells is used to analyze the natural circulation phenomenon. In the flow network analysis, the pressure drop for each assembly from the entrance of the fuel rack to the exit of the fuel assembly is balanced by the driving head due to the density difference between the pool fluid and the average fluid in each spent fuel assembly. The governing equations ore developed using this relation. But, since the parameters(flow rate, pressure loss coefficient, decay heat, density)are coupled each other, iteration method is used to obtain the solution. For the analysis of the YGN 3&4 spent fuel rack, 12 channels are considered and the inputs such as decay heat and pressure loss coefficient are determined conservatively. The results show the thermal-hydraulic characteristics(void fraction, density, boiling height)of the YGN 3&4 spent fuel rack. There occurs small amount of boiling in the cells. Fuel cladding temperature is lower than 343.3$^{\circ}C$. The evaluation of parametric effect indicates that flow resistances by geometric effect are very sensitive to Reynolds number in the transition region and the gap flow is negligible because of the larger flow resistance in the gap flow path than in the fuel bundle.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.214-225
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• 2003

A teleoperated manipulator system has been developed for remote handling of the spent fuel bundles. A heavy-duty power manipulator with high reduction ratio joints is used for the slave manipulator in the developed system since the handling tasks of the spent fuel bundles need power. Also, the universal type master manipulator, which has force reflecting capability, is used for precise remote manipulation. The power manipulators so frequently occur the control input saturation that the precise control performances are not achieved due to the windup phenomenon. An advanced bilateral control scheme compensating for the saturation is applied to the teleoperated manipulator system. The validity of the developed system is verified by the grid cutting and fuel transportation tasks from the mockup spent fuel bundle.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.11
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• pp.2986-2998
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• 1994

• Proceedings of the SAREK Conference
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• 2004.11a
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• pp.126-126
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• 2004