참고문헌
- Idaho National Laboratory, "Summary for the Next Generation Nuclear Plant Project in Review," INL/EXT-10-19142, Revision 1 (2010).
- J. Chang, Y. W. Kim, K. Y. Lee, Y. W. Lee, W. J. Lee, J. M. Noh, M. H. Kim, H. S. Lim, Y. J. Shin, K. K. Bae, and K. D. Jung, "A Study of a Nuclear Hydrogen Production Demonstration Plant," Nucl. Eng. Technol., vol. 39, no. 2, pp. 111-122 (2007). https://doi.org/10.5516/NET.2007.39.2.111
- J. M. Noh et al., "Development of Very High Temperature Reactor Design Technology," KAERI/RR-3462/2011 (Written in Korean), Korea Atomic Energy Research Institute (2012).
- C. K. Jo, H. S. Lim, and J. M. Noh, "Preconceptual Designs of the 200MWth Prism and Pebble-bed Type VHTR Cores," PHYSOR 2008, Interlaken, Switzerland, Sep. 14-19, 2008.
- A. J. Neylan, D. V. Graf, and A. C. Millunzi, "The Modular High Temperature Gas-Cooled Reactor (MHTGR) in the U.S.," Nucl. Eng. Des., vol. 109, pp. 99-105 (1988). https://doi.org/10.1016/0029-5493(88)90146-X
- S. Shiozawa, S. Fujikawa, T. Iyoku, K. Kunitomi, and Y. Tachibana, "Overview of HTTR design features," Nucl. Eng. Des., vol. 233, pp. 11-21 (2004). https://doi.org/10.1016/j.nucengdes.2004.07.016
- G. Melese and R. Katz, Thermal and Flow Design of Helium-Cooled Reactors, American Nuclear Society, Illinois USA (1984).
- S. Maruyama, N. Fujimoto, Y. Sudo, Y. Kiso, H. Hayakawa, "Fuel Temperature Analysis Method for Channel-Blockage Accident in HTTR," Nucl. Eng. Des., vol. 150, pp. 69-80 (1994). https://doi.org/10.1016/0029-5493(94)90052-3
- N. I. Tak, M. H. Kim, H. S. Lim, and J. M. Noh, "A Practical Method for Whole Core Thermal Analysis of Prismatic Gas-Cooled Reactor", Nucl. Tech., vol. 177, pp. 352-365 (2012). https://doi.org/10.13182/NT12-A13480
- ANSYS Inc., http://www.ansys.com (2014).
- CD-adapco, http://www.cd-adapco.com (2014).
- N. I. Tak, M. H. Kim, and W. J. Lee," Numerical Investigation of a Heat Transfer within the Prismatic Fuel Assembly of a Very High Temperature Reactor," Ann. Nucl. Energy, vol. 35, pp. 1892-1899 (2008). https://doi.org/10.1016/j.anucene.2008.04.005
- M. H. Kim, N. I. Tak, and J. M. Noh, "CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core," Transactions of the Korean Nuclear Society Autumn Meeting, Jeju, Korea, Oct. 21-22, 2010.
- O. Cioni, M. Marchand, G. Geffraye, and F. Ducros, "3D Thermal-Hydraulic Calculations of a Modular Block-type HTR Core," Nucl. Eng. Des., vol. 236, pp. 565-573 (2006). https://doi.org/10.1016/j.nucengdes.2005.10.024
- W. D. Pointer and J. W. Thomas, "Steady-State, Whole-Core Prismatic VHTR Simulation Including Core Bypass," Proceedings of ICAPP '10, San Diego, CA, USA, June 13-17, 2010, Paper 10310.
- W. J. Lee, J. J. Jeong, S.W. Lee and J. Chang, "Development of MARS-GCR/V1 for Thermal-Hydraulic Safety Analysis of Gas-Cooled Reactor Systems," Nucl. Eng. Technol., vol. 37, no. 6, pp. 587-594 (2005).
- H. S. Lim and H. C. NO, "GAMMA Multidimensional Multicomponent Mixture Analysis to Predict Air Ingress Phenomena in an HTGR," Nucl. Sci. Eng., vol. 152, pp. 87-97 (2006). https://doi.org/10.13182/NSE06-5
- C. L. Wheeler, C. W. Stewart, R. J. Cena, D. S. Rowe, and A. M. Sutey, "COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores," BNWL-1962, Battelle Pacific Northwest Lab. (1976).
- B. W. Travis and M. S. El-Genk, "Thermal-Hydraulics Analyses for 1/6 Prismatic VHTR Core and Fuel Element with and without Bypass Flow," Energy Conversion and Management, vol. 67, pp. 325-341 (2013). https://doi.org/10.1016/j.enconman.2012.11.003
- N. I. Tak , M. H. Kim, H. S. Lim, "Development of a Unit- Cell Based Multi-dimensional Heat Conduction Model for a Prismatic Fuel Block," Transactions of the Korean Nuclear Society Spring Meeting, Pyeongchang, Korea, May 27-28, 2010.
- H. K. Versteeg and W. Malalasekera, An Introduction to Computational Fluid Dynamics, The Finite Volume Method, Second Edition, Pearson Education Limited (2007).
- S.V. Patankar, Numerical Heat Transfer and Fluid Flow, Hemisphere, New York (1980).
- G. P. Greyvenstein and D. P. Laurie, "A Segregated CFD Approach to Pipe Network Analysis," Int. J. Numer. Meth. Engng, vol. 37, pp. 3685-3705 (1994). https://doi.org/10.1002/nme.1620372107
- G. P. Greyvenstein, "An Implicit Method for the Analysis of Transient Flows in Pipe Networks," Int. J. Numer. Meth. Engng, vol. 53, pp. 1127-1143 (2002). https://doi.org/10.1002/nme.323
- D. M. McEligot, G. E. McCreery, R. R. Schultz, J. Lee, P. Hejzlar, P. Stahle, P. Saha, "Investigation of Fundamental Thermal-Hydraulic Phenomena in Advanced Gas-Cooled Reactors," INL/EXT-06-11801, Idaho National Laboratory (2006).
- H. Kaburaki and T. Takizuka, "Effect of Crossflow on Flow Distribution in HTGR Core Column," J. Nucl. Sci. Tech., vol. 24 (7), pp. 516-525 (1987). https://doi.org/10.1080/18811248.1987.9735842
- N. I. Tak, M. H. Kim, and W. J. Lee, "A Benchmark CFD Calculation for a Cross Flow between Fuel Blocks of a Prismatic VHTR," Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 22, 2009.
- M. H. Kim, "CFD Model for Hot Spot Analysis of Standard Fuel Block," NHDD-RD-CA-11-005, Korea Atomic Energy Research Institute (2011).
- S. N. Lee, N. I. Tak, M. H. Kim, and J. M. Noh, "Thermo- Fluid Verification of Fuel Column with Crossflow Gap," Transactions of the Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, Oct. 24-25, 2013.
- N. I. Tak, M. H. Kim, and H. S. Lim, "Parallel Computation for Whole Core Thermo-fluid Simulation of Prismatic Gas-Cooled Reactor," Transactions of the Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, Oct. 27-28, 2011.
- C. K. Jo, Preliminary Result File of MASTER-GCR Calculation for PMR600, 12BH_B4C_080_R48_MAS_AS_PF3_Case27_summary.dat (2013).
-
Intel Corporation, "
$Intel^{(R)}$ MPI Library for Windows OS, Reference Manual," Rev. 4.0 Update 1, http://www.intel.com (2010). - J. Y. Cho et al., "DeCART v1.2 User's Manual," KAERI/TR-3438/2007, Korea Atomic Energy Research Institute (2007).