Nuclear Engineering and Technology

  • 제45권6호
  • /
  • Pages.745-752
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  • 2013
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

  • Tak, Nam-Il (Korea Atomic Energy Research Institute) ;
  • Kim, Min-Hwan (Korea Atomic Energy Research Institute) ;
  • Lim, Hong-Sik (Korea Atomic Energy Research Institute) ;
  • Noh, Jae Man (Korea Atomic Energy Research Institute) ;
  • Drzewiecki, Timothy J. (Department of Nuclear Engineering and Radiological Sciences, University of Michigan) ;
  • Seker, Volkan (Department of Nuclear Engineering and Radiological Sciences, University of Michigan) ;
  • Downar, Thomas J. (Department of Nuclear Engineering and Radiological Sciences, University of Michigan) ;
  • Kelly, Joseph (US Nuclear Regulatory Commission)
  • 투고 : 2013.09.24
  • 심사 : 2013.10.02
  • 발행 : 2013.11.25
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초록

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

키워드

참고문헌

  1. J. M. NOH et al., Development of Very High Temperature Reactor Design Technology, Korea Atomic Energy Research Institute (KAERI), KAERI/RR-3462/2011, (Written in Korean) (2012).
  2. M. H. KIM and H. S. LIM, "Evaluation of the Influence of Bypass Flow Gap Distribution on the Core Hot Spot in a Prismatic VHTR Core," Nucl. Eng. Des., 241, 3076 (2011). https://doi.org/10.1016/j.nucengdes.2011.05.009
  3. H. S. LIM and H. C. NO, "GAMMA multidimensional multicomponent mixture analysis to predict air ingress phenomena in an HTGR," Nucl. Sci. Eng., 152, 1 (2006).
  4. V. SEKER, Multiphysics Methods Development for High Temperature Gas Cooled Reactor, Ph.D Dissertation, Purdue University (2007).
  5. T. J. DRZEWIECKI, V. SEKER, J. M. KELLY, and T. J. DOWNAR, "The US NRC Advanced Gas REactor Evaluator (AGREE)," The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), Toronto, Ontario, Canada, Sep. 25-30 (2011).
  6. US NUCLEAR REGULATORY COMMISSION and US DEPARTMENT OF ENERGY, Next Generation Nuclear Plant Licensing Strategy. A Report to Congress (2008).
  7. S. J. YOON, Y. J. CHO, K. Y. KIM, M. H. KIM, and G. C. PARK, "Experimental Evaluation of the Bypass Flow in the VHTR Core," Transactions of SMiRT 19, Aug. 12- 17, 2007, Toronto, Ontario, CANADA, Paper #S03/5 (2007).
  8. G. C. PARK, S. J. YOON, and C. Y. JIN, "Fundamental Thermal-Fluid Experiment for Optimum Design of Very High Temperature Reactor," KAERI/CM-1127/2008, Apr. 2009 (Written in Korean) (2009).
  9. S. J. YOON, C. Y. JIN, M. H. KIM, and G. C. PARK, "Experimental and Computational Assessment of Core Bypass Flow in Block-Type Very High Temperature Reactor," Nucl. Technol., 175, 419 (2011). https://doi.org/10.13182/NT11-A12313
  10. S. J. YOON, C. Y. JIN, J. H. LEE, M. H. KIM, and G. C. PARK, "Study on the Flow Distribution in Prismatic VHTR Core with a Multi-Block Experiment and CFD Analysis," Nucl. Eng. Des., 241, 5174 (2011). https://doi.org/10.1016/j.nucengdes.2011.09.008
  11. G. C. PARK, S. J. YOON, J. H. LEE, and Y. G. LEE, "Fundamental Thermal-Fluid Experiments for the Evaluation of the Bypass Flow in the VHTR Core," KAERI/CM- 1449/2010, Dec. 2011 (Written in Korean) (2011).
  12. M. H. Kim et al., "Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor," Korea Atomic Energy Research Institute (KAERI), KAERI /RR-3293/2010, Dec. 2011 (Written in Korean) (2011).
  13. S. J. YOON, J. H. LEE, M. H. KIM, and G. C. PARK, "The Effects of Crossflow Gap and Axial Bypass Gap Distribution on the Flow Characteristics in Prismatic VHTR Core," Nucl. Eng. Des., 250, 465(2012). https://doi.org/10.1016/j.nucengdes.2012.04.025
  14. W. D. POINTER, "CFD Predictions of Gap Bypass in Prismatic VHTR Cores," Transactions of the American Nuclear Society, Nov. 7-11, 2010, Las Vegas, Nevada, VOL. 103, p. 897 (2010).
  15. ANSYS INC., http://www.ansys.com (2012).
  16. CD-ADAPCO, http://www.cd-adapco.com (2012).
  17. M. H. KIM, H. S. LIM, and N. I. TAK, "Computational Assessment of Bypass Flow in a Multi-Block Air Test," Transactions of the Korean Nuclear Society Spring Meeting, Pyeongchang, Korea, May 27-28 (2010).
  18. H. KABURAKI and T. TAKIZUKA, "Crossflow Characteristics of HTGR Fuel Blocks," Nucl. Eng. Des., 120, 425 (1990). https://doi.org/10.1016/0029-5493(90)90392-B