References
- J. KUPITZ, “Integration of Nuclear Energy and Desalination Systems,” Proc. Symp. Desalination of Seawater with Nuclear Energy, Daejeon, Korea, May 26–30, 1997, p. 20, IAEA-SM-347, International Atomic Energy Agency (1997).
- T. KUSUNOKI, N. ODANO, T. YORITSUNE, T. ISHIDA, T. HOSHI, and K. SAKO, “Design of Advanced Integral-Type Marine Reactor, MRX,” Nucl. Eng. Des., 201, 155 (2000).
- M. D. CARELLI, K. MILLER, C. V. LOMBARDI, N. E. TODREAS, E. GREENSPAN, H. NINOKATA, J. LOPEZ, L. CINOTTI, J. COLLADO, F. ORIOLO, G. ALONSO, M. M. MORAES, R. BOROUGHS, A. BARROSO, D. INGERSOLL, and N. CAVLINA, “IRIS: Proceeding Towards the Preliminary Design,” Proc. 10th Int. Conf. Nuclear Engineering (ICONE-10), Arlington, Virginia, April 14–18, 2002.
- M. H. CHANG, “Basic Design Report of SMART,” KAERI/TR-2142/2002, Korea Atomic Energy Research Institute (2002).
- S.-H. KIM, K. K. KIM, J. W. YEO, M. H. CHANG, and S. Q. ZEE, “Design Verification Program of SMART,” Proc. GENES4/ANP2003, Kyoto, Japan, September 15–19, 2003.
- H. Y. YOON, “Thermal Hydraulic Model Description of TASS/SMR,” KAERI/TR-1835/2001, Korea Atomic Energy Research Institute (2001).
- S. SIM, “TASS Code Technical Report: Volume 1 TASS Code Technical Manual,” KAERI/TR-845/97, Korea Atomic Energy Research Institute (1997).
- RELAP5 DEVELOPMENT TEAM, “RELAP5 Code Manual,” NUREG/CR-5535, Idaho National Engineering Laboratory (1998).
- S. SIM, “TASS Code Technical Report: Volume 2 TASS Code Validation Report for the Non-LOCA Transient Analysis of the CE and Westinghouse Type Plants,” KAERI/TR-845-1/97, Korea Atomic Energy Research Institute (1997).
- V. H. GLAHN, “An Empirical Relation for Predicting Void Fraction with Two-Phase Steam Water Flow,” NASATechnical Note D-1189 (1962).
- B. CHEXAL, G. LELLOUCHE, J. HOROWITZ, and J. HEALZER, “A Void Fraction Correlation for Generalized Applications,” Prog. Nucl. Energy, 27, 4, 255 (1992).
- D. H. HWANG, G. W. SEO, J. C. LEE, and K. K. KIM, “Development of CHF Correlation Systems for SMART-P Fuel Assembly,” KAERI/TR-2943/2005, Korea Atomic Energy Research Institute (2005).
- D. H. HWANG, “Development of a Thermal Hydraulic Design Methodology for an Advanced Reactor Core with Vertical Parallel Channels,” KAERI/TR-992/98, Korea Atomic Energy Research Institute (1998).
- Y. MORI and W. NAKAYAMA, “Study on Forced Convective Heat Transfer in Curved Pipes (2nd Report, Turbulent Region),” Int. J. Heat Mass Transfer, 10, 37 (1967).
- S. W. CHURCHILL and H. H. S. CHU, “Correlating Equations for Laminar and Turbulent Free Convection from a Horizontal Cylinder,” Int. J. Heat Mass Transfer, 18, 1049 (1975).
- A. ZUKAUSKAS, “Heat Transfer from Tubes in Crossflow,” Advanced Heat Transfer, Vol. 8, p. 93, Academic Press (1972).