Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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A MIXED CORE FOR SUPERCRITICAL WATER-COOLED REACTORS

  • Cheng, Xu (School of Nuclear Science and Engineering, Shanghai Jiao Tong University) ;
  • Liu, Xiao-Jing (School of Nuclear Science and Engineering, Shanghai Jiao Tong University) ;
  • Yang, Yan-Hua (School of Nuclear Science and Engineering, Shanghai Jiao Tong University)
  • Published : 2008.03.31
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Abstract

In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor(SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.

Keywords

References

  1. W.Q. Shen, 'Strategy Arrangement of Nuclear Power in China', 6th China-European Energy Co-operation Conference, Shanghai, China, Feb. 20-21, 2006
  2. Y. Oka, 'Review of high temperature water and steam cooled reactor concepts', Proc. of SCR-2000, Tokyo, Japan, Nov.6-8, 2000
  3. D. Squarera, T. Schulenberg, D. Struwea, Y. Oka et al. 'High Performance Light Water Reactor', Nuclear Engineering and Design, 221, (2003)
  4. P. McDonald, J. Buongiorno, J.W. Sterbentz, C. Davis, R. Witt, 'Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production', INEEL/EXT-04- 02530, INEEL, January, 2005
  5. A.Yamaji, K. Kaemi, Y. Oka, S. Koshizuka et al. 'Improved Core Design of High Temperature Supercritical-Pressure Light Water Reactor', Proc. of ICAPP'04, Pittsburgh, Pa, USA, June 13-17, 2004
  6. Y.Y. Bae, H.K. Joo, J. Jang, J. Song, H.Y. Yoon, 'Research of a Supercritical Pressure Water Cooled Reactor in Korea', Proc. of ICAPP'04, Pittsburgh, Pa, USA, June 13-17, 2004
  7. S.J. Bushby, G.R. Dimmick, R.B. Duffey, 'Conceptual Designs for Advanced, High-Temperature CANDU Reactors', Proc. of SCR-2000, Tokyo, Japan, Nov.6-8, 2004
  8. X. Cheng, T. Schulenberg, D. Bittermann, P. Rau, 'Design Analysis of Core Assemblies for Supercritical Pressure Conditions', Nuclear Engineering and Design, 223, 3, (2003)
  9. J. Hofmeister, E. Laurien, A.G. Class, T. Schulenberg, 'Turbulent Mixing in the Foot Piece of a HPLWR Fuel Assembly', GLOBAL 2005, Tuskuba, Japan, Oct. 9-13, 2005
  10. J. Bourngiorno, Private Communication, 2002
  11. W. Oldekop, H.D. Berger, W. Zeggel, 'General Features of Advanced Pressurized Water Reactors with Improved Fuel Utilization', Nuclear Technology, 59(1982)
  12. X. Cheng, T. Schulenberg, 'Heat Transfer at Supercritical Pressures - Literature Review and Application to a HPLWR', Scientific Report FZKA6609, Forschungszentrum Karlsruhe, May 2001
  13. X. J. Liu, Y.H. Yang, X. Cheng, 'Design Analysis of a SCWR Fuel Assembly Using a Coupled Method', 3rd int. Symposium on Supercritical Water-Cooled Reactors - Design and Technology, Shanghai, China, March 12-15, 2007
  14. A.A. Bishop, R.O. Sandberg, L.S. Tong, 'Forced convection heat transfer to water at near critical temperatures and supercritical pressures', WCAP-5549-P, Part-III-B,1964

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