95
Views
1
CrossRef citations to date
0
Altmetric
Technical Paper

Thermal Hydraulics of Sodium-Cooled Fast Reactors: Key Design and Safety Issues and Highlights

&
Pages 11-23 | Published online: 10 Aug 2017

References

  • S. Kotake et al., “Development of Advanced Loop-Type Fast Reactor in Japan,” Nucl. Technol., 170, 133 (2010).
  • K. Ishii et al., “Activities for 4S USNRC Licensing,” Prog. Nucl. Energy, 53, 7, 831 (2011); see also http://pbadupws.nrc.gov/docs/ML0814/ —ML081440765: “Design Description for the 4S Reactor” (May 2008); ML082050556: “Long-Life Metallic Fuel for the Super Safe, Small and Simple (4S) Reactor” (June 2008); ML092170507: “4S Safety Analysis” (July 2009); ML101400662: “Phenomena Identification and Ranking Tables (PIRTs) for the 4S and Further Investigation Program—Loss of Offsite Power, Sodium Leakage from Intermediate Piping, and Failure of a Cavity Can Events” (May 2010).
  • S. Nakanishi et al., “Development of Passive Shutdown System for SFR,” Nucl. Technol., 170, 181 (2010).
  • M. Takamatsu et al., “Demonstration of Control Rod Holding Stability of the Self Actuated Shutdown System in Joyo for Enhancement of Fast Reactor Inherent Safety,” J. Nucl. Sci. Technol, 44, 511 (2007).
  • H. Yamano et al., “Development of Advanced Loop-Type Fast Reactor in Japan (2): Technical Feasibility of Two-Loop Cooling System in JSFR,” Proc. Int. Congress Advances in Nuclear Power Plants (ICAPP’08-8231), Anaheim, California, June 8–12, 2008.
  • Y. Tobita et al., “Analytical Study on Elimination of Severe Recriticalities in Large Scale LMFBRS with Enhancement of Fuel Discharge,” Nucl. Eng. Des., 238, 57 (2008).
  • N. Uto et al., “Conceptual Design for Japan Sodium-Cooled Fast Reactor (1): Current Status of System Design for JSFR,” Proc. Int. Congress Advances in Nuclear Power Plants (ICAPP’09-9298), Tokyo, Japan, May 10–14, 2009.
  • S. Okamura et al., “Seismic Isolation Design for JSFR,” Proc. Int. Conf. Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, Kyoto, Japan, 2009, IAEA-CN-176-08-28P.
  • M. Tabuchi and Y. Takahashi, “Evaluation of Creep Strength Reduction Factors for Welded Joints of Modified 9Cr-1Mo Steel (P91),” Proc. ASME Pressure Vessels and Piping Division Conf. (PVP2006-ICPVT-11-93350), Vancouver, British Columbia, Canada, July 23–27, 2006.
  • D. Pialla et al., “Natural Convection Test in Phénix Reactor and Associated CATHARE Calculation,” Proc. 14th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH14-099), Toronto, Canada, September 25–30, 2011.
  • A. Chenu et al., “One- and Two-Dimensional Simulations of Sodium Boiling Under Loss-of-Flow Conditions in a Pin Bundle with the TRACE Code,” Proc. 13th Int. Topl. Mtg. Nuclear Reactor Thermalhydraulics (N13P1108), Kanzawa, Japan, September 28–October 2, 2009.
  • M. Z. Podowski et al., “Multidimensional Analysis of Fission Gas Discharge Following Fuel Element Failure in Sodium Fast Reactor,” Proc. 13thInt. Topl. Mtg. Nuclear Reactor Thermalhydraulics (N13P1294), Kanzawa, Japan, September 28–October 2, 2009.
  • I. Banerjee et al., “Experimental Assessment of Control Plug Hydraulics in PFBR,” Proc. 13th Int. Topl. Mtg. Nuclear Reactor Thermalhydraulics (N13P1120), Kanzawa, Japan, September 28–October 2, 2009.
  • Japan Atomic Energy Agency Web Site: http://www.jaea.go.jp/04/monju/EnglishSite/index.en.html.
  • “Toshiba (Slide Presentations), 4S Reactor—Super-Safe, Small and Simple—Meetings with NRC Pre-Application Review” —ML072950025 (Oct. 22,2007); ML080600037 (Feb. 21,2008); ML081400095 (May 21,2008); ML082190834 (Aug. 8, 2008); available at U.S. Nuclear Regulatory Commission Web Site (2007–2008): http://pbadupws.nrc.gov/docs/.
  • H. Ninokata, E. Merzari, and A. Khakim, “Analysis of Low Reynolds Number Turbulent Flow Phenomena in Nuclear Fuel Pin Subassemblies of Tight Lattice Configuration,” Nucl. Eng. Des., 239, 855 (2009).
  • E. Merzari et al., “Biglobal Linear Stability Analysis for the Flow in Eccentric Annular Channels and a Related Geometry,” Phys. Fluids, 20, 114104, doi:10.1063/1.3005864 (Nov. 25, 2008).
  • N. Kimura et al., “Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor,” Nucl. Technol., 152, 210 (2005).
  • ANSI/HI9.8, “American National Standard for Pump Intake Design,” American National Standards Institute (1998).
  • TSJ S 002, “Standard Method for Model Testing the Performance of a Pump Sump,” Turbomachinery Society of Japan (2005).
  • T. Ezure et al., “Experimental Study on Influences of Kinematic Viscosity on Occurrences of Cavitation due to SubSurface Vortex,” Proc. 14th Int. Topl. Mtg. Nuclear Reactor Thermalhydraulics (NURETH14-276), Toronto, Canada, September 25–30, 2011.
  • A. Ono et al., “Influence of Elbow Curvature on Flow Structure at Elbow Outlet Under High Reynolds Number Condition,” Nucl. Eng. Des., 241, 4409 (2011).
  • H. Kamide et al., “Sodium Experiment on Fully Natural Circulation Systems for Decay Heat Removal in Japan Sodium-Cooled Fast Reactor,” Proc. 14th Int. Topl. Mtg. Nuclear Reactor Thermalhydraulics (NURETH14-179), Toronto, Canada, September 25–30, 2011.
  • H. Yamano, Y. Tobita, and I. Sato, “Development of Technical Database in the Unprotected Events for Level 2 PSA of Sodium-Cooled Fast Reactors,” Proc. 7th Japan-Korea Symp. Nuclear Thermal Hydraulics and Safety (NTHAS7), Chun-cheon, Korea, November 14–17, 2010.
  • K. Konishi et al., “Progress in Establishment of the Innovative Safety Logic for SFR Eliminating the Recriticality Issue with the EAGLE Experimental Program,” presented at Int. Scientific-Practical Conf. Nuclear Power Engineering in Kazakhstan, Kurchatov, Kazakhstan, June 11–13, 2008.
  • K. Aoto et al., “Design Study and R&D Progress on Japan Sodium-Cooled Fast Reactor,” J. Nucl. Sci. Technol., 48, 4, 463 (2011).
  • Y. Tobita et al., “Development of Severe Accident Evaluation Technology (Level 2 PSA) for Sodium-Cooled Fast Reactors—(3) Identification of Dominant Factors in Transition Phase of Unprotected Events,” Proc. Int. Congress Advances in Nuclear Power Plants (ICAPP’09-9127), Tokyo, Japan, May 10–14, 2009.
  • J. Toyooka et al., “SIMMER-III Analysis of EAGLE-1 In-Pile Tests Focusing on Heat Transfer from Molten Core Material to Steel-Wall Structure,” Proc. 7th Japan-Korea Symp. Nuclear Thermal Hydraulics and Safety (NTHAS7), Chun-cheon, Korea, November 14–17, 2010.

Reprints and Corporate Permissions

Please note: Selecting permissions does not provide access to the full text of the article, please see our help page How do I view content?

To request a reprint or corporate permissions for this article, please click on the relevant link below:

Academic Permissions

Please note: Selecting permissions does not provide access to the full text of the article, please see our help page How do I view content?

Obtain permissions instantly via Rightslink by clicking on the button below:

If you are unable to obtain permissions via Rightslink, please complete and submit this Permissions form. For more information, please visit our Permissions help page.