Assessment of the local hot core temperature during normal operation in a pebble-bed type of Very High Temperature Reactor (VHTR) has been carried out by using the Computational Fluid Dynamic (CFD) method for which the boundary conditions were obtained from the results of a macroscopic analysis of the core using a system thermal analysis code, GAMMA. Three pebble arrangements are selected, which are Simple Cubic (SC), Body-Centered Cubic (BCC), and Face-Centered Cubic (FCC). Results showed that the SC arrangement having the lowest porosity gives the highest fuel temperature of 1237°C but still below the normal operational fuel limit of 1250°C. Comparison of the CFD results with an empirical correlation was made for the pressure drop and the Nusselt number but there were large differences between them. The benchmark calculation of a pressure drop for packed particles in a square channel indicated that the correlation for the full core used in the system code is not appropriate for the prediction of a local thermal fluid behavior.
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16th International Conference on Nuclear Engineering
May 11–15, 2008
Orlando, Florida, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-4815-9
PROCEEDINGS PAPER
CFD Assessment of the Local Hot Core Temperature in a Pebble-Bed Type Very High Temperature Reactor
Min-Hwan Kim,
Min-Hwan Kim
Korea Atomic Energy Research Institute, Daejeon, South Korea
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Hong-Sik Lim,
Hong-Sik Lim
Korea Atomic Energy Research Institute, Daejeon, South Korea
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Won Jae Lee
Won Jae Lee
Korea Atomic Energy Research Institute, Daejeon, South Korea
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Min-Hwan Kim
Korea Atomic Energy Research Institute, Daejeon, South Korea
Hong-Sik Lim
Korea Atomic Energy Research Institute, Daejeon, South Korea
Won Jae Lee
Korea Atomic Energy Research Institute, Daejeon, South Korea
Paper No:
ICONE16-48314, pp. 281-287; 7 pages
Published Online:
June 24, 2009
Citation
Kim, M, Lim, H, & Lee, WJ. "CFD Assessment of the Local Hot Core Temperature in a Pebble-Bed Type Very High Temperature Reactor." Proceedings of the 16th International Conference on Nuclear Engineering. Volume 2: Fuel Cycle and High Level Waste Management; Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition. Orlando, Florida, USA. May 11–15, 2008. pp. 281-287. ASME. https://doi.org/10.1115/ICONE16-48314
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