System Design and Analysis of a 900-MW(thermal) Lead-Cooled Fast Reactor

References

• “Handbook on Lead-Bismuth Eutetic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies,” Chap. 1, Organisation for Economic Co-operation and Development/Nuclear Energy Agency, Paris (2007).
   Google Scholar
• V. V. ORLOV et al., “Deterministic Safety of BREST Reactors,” presented at 11th Int. Conf. Nuclear Engineering (ICONE11), Tokyo, Japan, April 20–23, 2003.
   Google Scholar
• E. L. GLUEKLER, “U.S. Advanced Liquid Metal Reactor (ALMR),” Prog. Nucl. Energy, 31, 1/2, 43 (1997).
   Google Scholar
• J. H. EOH, E. K. KIM, and S. O. KIM, “A Study on the Characteristics of the Decay Heat Removal Capacity for a Large Thermal Rated LMR Design,” Proc. KNS Spring Mtg., Kyeongju, Korea, May 29–30, 2003, Korean Nuclear Society (2003).
   Google Scholar
• “Safety Related Terms for Advanced Nuclear Plants,” IAEATECDOC-626, International Atomic Energy Agency (Sep. 1991).
   Google Scholar
• I. J. KARASSIK, J. P. MESSINA, P. C. COOPER, and C. C. HEALD, Pump Handbook, 3rd ed., McGraw-Hill, New York (2001).
   Google Scholar
• V. Y. ABRAMOV, S. N. BOZIN, S. V. EVROPIN, B. S. RODCHENKOV, V. N. LEONOV, A. I. FILIN, and V. G. MARKOV, “Corrosion and Mechanical Properties of BREST-OD-300 Reactor Structural Materials,” presented at 11th Int. Conf. Nuclear Engineering (ICONE11), Tokyo, Japan, April 20–23, 2003.
   Google Scholar
• V. S. STEPANOV, Y. G. DRAGOONOV, V. M. KOUTANOV, V. A. SHOULYNDIN, B. F. GROMOV, and G. I. TOSHINSKY, “Steam Supply Unit BM-40/A Experience of Development and Exploitation,” Proc. Int. Conf. Heavy Liquid-Metal Coolants in Nuclear Technology, Obninsk, Russian Federation, October 5–9, 1998, Vol. 1 (1998).
   Google Scholar
• F. M. MITENKOV, G. M. ANTONOVSKY, A. A. BELJAEV, B. I. RUNOV, and M. V. SMIMOV, “Experience in Construction and Operation of OK-550 RP Equipment,” Proc. Int. Conf. Heavy Liquid-Metal Coolants in Nuclear Technology, Obninsk, Russian Federation, October 5–9, 1998, Vol. 1 (1998).
   Google Scholar
• “ANSYS Users Manual, Version 6.1,” ANSYS Inc. (2002).
   Google Scholar
• “ASME Boiler and PressureVessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Div. 1, Subsection NH, Class 1 Components in Elevated Temperature Service,” ASME (2004).
   Google Scholar
• Y. KIM, C. H. CHO, S. J. KIM, and T. Y. SONG, “Self-Sustaining Lead-Cooled Reactor with Single Fuel Enrichment,” Trans. Am. Nucl. Soc., 89, 645 (2003).
   Google Scholar
• A. G. CROFF, “A User’s Manual for the ORIGEN2 Computer Code,” ORNL/TM-7175, Oak Ridge National Laboratory (1980).
   Google Scholar
• B. J. TOPEL, “AUser’s Guide to the REBUS-3 Fuel Cycle Analysis Capability,” ANL-83-2, Argonne National Laboratory (1983).
   Google Scholar
• K. L. DERSTINE, “DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite Difference Diffusion Theory Problems,” ANL-82-64, Argonne National Laboratory (1984).
   Google Scholar
• R. E. MacFARLANE, “TRANSX2: A Code for Interfacing MATXS Cross Section Libraries to Nuclear Transport Codes,” LA-12312-MS, Los Alamos National Laboratory (1992).
   Google Scholar
• R. E. ALCOUFFE, F. W. BRINKLEY, JR., D. R. MARR, and R. D. O’DELL, “User’s Guide for TWODANT: A Code Package for Two-Dimensional, Diffusion-Accelerated Neutral Particle Transport,” LA-10049-M, Los Alamos National Laboratory (1990).
   Google Scholar
• P. A. FOMITCHENKO, “Physics of Lead-Cooled Reactors,” Proc. Frédéric Joliot Summer School in Reactor Physics, Cadarache, France, August 17–26, 1998 (1998).
   Google Scholar
• P. HEJZLAR, J. BUONGIORNO, P. E. MacDONALD, and N. E. TODREAS, “Design Strategy and Constraints for Medium-Power Lead-Alloy-Cooled Actinide Burners,” Nucl. Technol., 147, 321 (2004).
   Web of Science ®Google Scholar
• “ANSYS CFXVersion 5.7.1,” ANSYS Europe, Ltd. (2004).
   Google Scholar
• C. G. SPEZIALE, S. SARKAR, and T. B. GATSKI, “Modeling the Pressure-Strain Correlation of Turbulence: An Invariant Dynamical Systems Approach,” J. Fluid Mech., 277, 245 (1991).
   Google Scholar
• A. O. DEMUREN and W. RODI, “Calculation of Turbulent Driven Secondary Motion in Non-Circular Duct,” J. Fluid Mech., 140, 189 (1989).
   Google Scholar
• Y. M. KWON, Y. B. LEE, W. P. CHANG, and D. HAHN, “SSC-K Code User’s Manual, Rev. 1,” KAERI/TR-2014/2002, Korea Atomic Energy Research Institute (2002).
   Google Scholar
• K. S. HA, W. P. CHANG, S. J. KIM, and Y. B. LEE, “Preliminary Safety Analysis for a Lead-Cooled Fast Reactor,” Proc. Int. Congress Advances in Nuclear Power Plants (ICAPP’05), Seoul, Korea, May 15–19, 2005 (2005).
   Google Scholar
• J. H. EOH, S. KIM, S. J. KIM, and S. O. KIM, “Feasibility Study of a Passive DHR System with Heat Transfer Enhancement Mechanism in a Lead-Cooled Fast Reactor,” Nucl. Technol., 160, 216 (2007).
   Web of Science ®Google Scholar
• Y. S. SIM, M. H. WI, and S. O. KIM, “Analysis on Decay Heat Removal Characteristics of PSDRS,” Proc. KNS Spring Conf., Suwon, Korea, May 1998, p. 653, Korean Nuclear Society (1998).
   Google Scholar