Core Design

  • Yoshiaki Oka
  • Seiichi Koshizuka
  • Yuki Ishiwatari
  • Akifumi Yamaji


This chapter presents the design concepts of the Super LWR core, including fuel rod and fuel assembly designs. The design concepts are comprehensively described with the design targets, criteria, margins, boundary conditions, and future requirements for further research and development. The design methods developed for the core designs are essentially equivalent to those methods adopted in designing commercial LWR cores. They include three-dimensional neutronic and thermal-hydraulic coupled core calculations, subchannel analyses, statistical design uncertainty evaluations based on a Monte Carlo sampling technique, and fuel rod behavior analyses. These analyses and the methods for developing the core concepts are described in detail. The evolutions of the core design concepts with new ideas and advances in the design methods are also described. The established core concept achieves an average core outlet temperature of 500°C at 25 MPa with a once-through direct cycle plant system.


Fuel Assembly Coolant Flow Rate Core Design Engineering Uncertainty Pseudocritical Temperature 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.




Auxiliary feedwater system


Anticipated transients without scram


Average linear heat generation rate


Beginning of cycle


Balance of plant


Beginning of life


Boiling water reactor


Computational fluid dynamics


Critical heat flux


Departure from nucleate boiling


Departure from nucleate boiling ratio


End of life


Fast reactor


Fast breeder reactor


Fuel cladding mechanical interaction


Fossil fired power plant


Fission product


Heat transfer deterioration


Liquid metal fast breeder reactor


Improved Thermal Design Procedure


Low leakage loading pattern


Loss of coolant accident


Light water reactor


Maximum cladding surface temperature


Monte Carlo Statistical Thermal Design Procedure


Minimum deterioration heat flux ratio


Optimized Monte Carlo Thermal Design Process


General Statistical Method


Maximum linear heat generation rate


Oxide dispersion strengthened


Pellet-cladding mechanical interaction


Pellet cladding interaction


Collision probability calculation module


Pressurized water reactor


Reactivity insertion accident


Reactor pressure vessel


Revised Thermal Design Procedure


Root Sum Square


Stress corrosion cracking


Statistical thermal design procedure


  1. 1.
    H. S. Swenson, J. R. Carver and C. R. Kakarala, “Heat transfer to supercritical water in smooth-bore tubes,” Journal of Heat Transfer, Vol. 87, 477–484 (1965)CrossRefGoogle Scholar
  2. 2.
    B. S. Shiralkar and P. Griffith, “Deterioration in heat transfer to fluids at supercritical pressure and high heat fluxes,” Journal of Heat Transfer, Vol. 91, 27–36 (1969)CrossRefGoogle Scholar
  3. 3.
    E. Stewart, P. Stewart and A. Watson, “Thermo-acoustic oscillations in forced convection heat transfer to supercritical pressure water,” International Journal of Heat and Mass Transfer, Vol. 16, 257–270 (1973)CrossRefGoogle Scholar
  4. 4.
    J. D. Jackson and W. B. Hall, Forced convection heat transfer to fluids at supercritical pressure. In Turbulent Forced Convection in Channels and Bundles, Vol. 2, 563–611. Hemisphere, New York (1979).Google Scholar
  5. 5.
    S. Koshizuka, N. Takano and Y. Oka, “Numerical analysis of deterioration phenomena in heat transfer to supercritical water,” International Journal of Heat and Mass Transfer, Vol. 38(16), 3077–3084 (1995)CrossRefGoogle Scholar
  6. 6.
    S. Koshizuka and Y. Oka, “Computational Analysis of Deterioration Phenomena and Thermal-Hydraulic Design of SCR,” Proc. SCR2000, Tokyo, November 6–8 (2000)Google Scholar
  7. 7.
    K. Kitoh, S. Koshizuka and Y. Oka, “Pressure and flow-induced accident and transient analyses of a direct-cycle, supercritical-pressure, light-water-cooled fast reactor,” Nuclear Technology, Vol. 123, 233–244 (1998)Google Scholar
  8. 8.
    K. Yamagata, K. Nishikawa, S. Hasegawa, T. Fujii and S. Yoshida, “Forced convection heat transfer to supercritical water flowing in tubes,” International Journal of Heat and Mass Transfer, Vol. 15, 2575–2593 (1972)CrossRefGoogle Scholar
  9. 9.
    A. Yamaji, “Fuel and Core Design of Super LWR,” Doctoral thesis, the University of Tokyo (2005) (in Japanese)Google Scholar
  10. 10.
    I. L. Pioro and R. B. Duffey, “Experimental heat transfer in supercritical water flowing inside channels (survey),” Nuclear Engineering and Design, Vol. 235, 2407–2430 (2005)CrossRefGoogle Scholar
  11. 11.
    H. Mori, S. Yoshida, et al., “Heat Transfer Study Under Supercritical Pressure Conditions for Single Rod Test Section,” Proc. ICAPP’05, Seoul, Korea, May 15–19, 2005, Paper 5303 (2005)Google Scholar
  12. 12.
    Y. Oka and S. Koshizuka, “Supercritical-pressure, once-through cycle light water cooled reactor concept,” Journal of Nuclear Science and Technology, Vol. 38(12), 1081–1089 (2001)CrossRefGoogle Scholar
  13. 13.
    Y. Oka, S. Koshizuka, Y. Ishiwatari, et al., “Overview of Design Studies of High Temperature Reactor Cooled by Supercritical Light Water at the University of Tokyo,” Proc. GENES4/ANP2003, Kyoto, Japan, September 15–19, Paper 1068 (2003)Google Scholar
  14. 14.
    K. Okumura, T. Kugo, K. Kaneko and K. Tsuchihashi “SRAC (Ver.2002); The comprehensive neutronics calculation code system,” Japan Atomic Energy Research Institute (JAERI) Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, JapanGoogle Scholar
  15. 15.
    K. Shibata, T. Kawano, T. Nakagawa, et al., “Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3,” Journal of Nuclear Science and Technology, Vol. 39(11), 1125–1136 (2002)CrossRefGoogle Scholar
  16. 16.
    “An evaluation of steam-cooled fast breeder reactors,” WASH-1088, 1969Google Scholar
  17. 17.
    M. Kureta, H. Tamai, A. Ohnuki, T. Sato, W. Liu and H. Akimito, “Critical power experiment with a Tight-Lattice 37-Rod Bundle,” Journal of Nuclear Science and Technology, Vol. 43(2), 198–205 (2006)CrossRefGoogle Scholar
  18. 18.
    K. Dobashi, “Conceptual Design of Supercritical-pressure Light Water Cooled and Moderated Reactor,” Doctoral thesis, the University of Tokyo (1998) (in Japanese)Google Scholar
  19. 19.
    Y. Oka, S. Koshizuka and T. Yamasaki, “Direct cycle light water reactor operating at supercritical pressure,” Journal of Nuclear Science and Technology, Vol. 29(6), 585–588 (1992)CrossRefGoogle Scholar
  20. 20.
    Y. Okano, S. Koshizuka and Y. Oka, “Core design of a direct-cycle, supercritical pressure, light water reactor with double tube water rods,” Journal of Nuclear Science and Technology, Vol. 33(4), 365–373 (1996)CrossRefGoogle Scholar
  21. 21.
    K. Kamei, “Core design of Super LWR and its safety analysis at subcritical-pressure,” Master’s thesis, the University of Tokyo (2006) (in Japanese)Google Scholar
  22. 22.
    K. Kamei, Y. Yamaji, Y. Ishiwatari, et al., “Fuel and core design of super light water reactor with low leakage fuel loading pattern,” Journal of Nuclear Science and Technology, Vol. 43(2), 129–139 (2006)CrossRefGoogle Scholar
  23. 23.
    T. Tanabe, S. Koshizuka and Y. Oka, “A Subchannel Analysis Code for Supercritical-Pressure LWR with Downward-Flowing Water Rods,” Proc. ICAPP’04, Pittsburgh, PA, June 13–17, 2004, Paper 4333 (2004)Google Scholar
  24. 24.
    T. Tanabe, “Subchannel Analysis of Super Light Water Reactor,” M.S. Thesis, University of Tokyo (2005) (In Japanese)Google Scholar
  25. 25.
    M. J. Watts and C. T. Chou, “Mixed Convection Heat Transfer to Supercritical Pressure Water,” Proc. 7th Int. Heat Transfer Conf., Munich, W. Germany, September 6–10, 1982, 495–500 (1982)Google Scholar
  26. 26.
    N. E. Todreas, and M. S. Kazimi, “Nuclear Systems I; Thermal Hydraulic Fundamentals,” Taylor & Francis, New York (1990)Google Scholar
  27. 27.
    J. Yang, Y. Oka, J. Liu, Y. Ishiwatari and A. Yamaji, “Development on statistical thermal design procedure to evaluate engineering uncertainty of super LWR,” Journal of Nuclear Science and Technology, Vol. 43(1), 32–42 (2006)CrossRefGoogle Scholar
  28. 28.
    S. Ray, A. J. Friedland and E. H. Novendstern, “Westinghouse Advanced Statistical DNB Methodology – The Revised Thermal Design Procedure,” Proc. NUTHOS-3, Seoul, Korea, November, 1988, A5-261 (1988)Google Scholar
  29. 29.
    L. S. Tong and J. Weisman, “Thermal Analysis of Pressurized Water Reactors 3rd Edition,” America Nuclear Society, 582 (1996)Google Scholar
  30. 30.
    N. E. Todreas and M. S. Kazimi, “Nuclear Systems II; Elements of Thermal Hydraulic Design,” Hemisphere Publ. Corp., New York (1990)Google Scholar
  31. 31.
    J. Robeyns, F. Parmentier and G. Peeters, “Application of a Statistical Thermal Design Procedure to Evaluate the PWR DNBR Safety Analysis Limits,” Proc. ICONE 9, Nice, France, April 8–12, 2001, ICONE-9091 (2001)Google Scholar
  32. 32.
    J. P. Bourteele, J. Greige and M. Missaglia, “The Framatome Generalized Statistical DNBR Method (MSG),” Proc. NURETH-6, Grenoble, France, October 5–8, 1993, v.1-355 (1993)Google Scholar
  33. 33.
    K. L. Eeckhout and J. J. Robeyns, “MTDP – An Optimized MONTE CARLO Method for Evaluation of the PWR Core Thermal Design Margin,” Proc. NURETH-8, Kyoto, Japan, September 30–October 4, 1997, v.1-421 (1997)Google Scholar
  34. 34.
    A. Yamaji, Y. Oka, J. Yang, J. Liu, Y. Ishiwatari and S. Koshizuka, “Design and Integrity Analyses of the Super LWR Fuel Rod,” Proc. Global 2005, Tsukuba, Japan, October 9–13, 2005 Paper 556 (2005)Google Scholar
  35. 35.
    M. Suzuki and H. Saitou, Light Water Reactor Fuel Analysis Code FEMAXI-6(Ver.1), JAEA Data/Code 2005-003 (2005)Google Scholar

Copyright information

© Springer Science+Business Media, LLC 2010

Authors and Affiliations

  • Yoshiaki Oka
    • 1
  • Seiichi Koshizuka
    • 2
  • Yuki Ishiwatari
    • 3
  • Akifumi Yamaji
    • 4
  1. 1.Department of Nuclear Energy Graduate School of Advanced Science and EngineeringWaseda UniversityShinjuku-kuJapan
  2. 2.Department of Systems Innovation Graduate School of EngineeringUniversity of TokyoBunkyo-kuJapan
  3. 3.Department of Nuclear Engineering and Management Graduate School of EngineeringUniversity of TokyoBunkyo-kuJapan
  4. 4.Department of Nuclear Engineering and ManagementUniversity of TokyoBunkyo-kuJapan

Personalised recommendations