In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based turbulence models for the analysis of flow distribution inside the 1/5 scaled-down APR+ (Advanced Power Reactor Plus), the simulation was conducted with the commercial computational fluid dynamics software, ANSYS CFX R.13. The results predicted were then compared with the measured data. It was concluded that reactor internal-flow pattern differed locally; depending on the turbulence models used. In particular, the prediction performance of turbulence models showed the largest difference in the regions from the flow skirt to fuel assembly inlet. The prediction performance of the k-ε model was superior to other turbulence models. This model also predicted the relatively uniform distribution of core-inlet flow-rate.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4591-2
PROCEEDINGS PAPER
Sensitivity Study on Turbulence Models for the Prediction of the Reactor Internal Flow
Gong Hee Lee,
Gong Hee Lee
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
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Young Seok Bang,
Young Seok Bang
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
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Sweng Woong Woo,
Sweng Woong Woo
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
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Ae Ju Cheong
Ae Ju Cheong
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
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Gong Hee Lee
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
Young Seok Bang
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
Sweng Woong Woo
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
Ae Ju Cheong
Korea Institute of Nuclear Safety, Daejeon, Republic of Korea
Paper No:
ICONE22-31255, V02BT09A064; 8 pages
Published Online:
November 17, 2014
Citation
Lee, GH, Bang, YS, Woo, SW, & Cheong, AJ. "Sensitivity Study on Turbulence Models for the Prediction of the Reactor Internal Flow." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2B: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02BT09A064. ASME. https://doi.org/10.1115/ICONE22-31255
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