Core and Fuel Design (From Mechanism to Structure)

  • Nobuo NakaeEmail author
  • Toshikazu Takeda
  • Hiroyuki Ohshima
Part of the An Advanced Course in Nuclear Engineering book series (ACNE, volume 8)


This chapter introduces the procedure and basic concept of nuclear design, thermal hydraulic design, and plant dynamics analysis. The outline of reactor physics and thermal hydraulic engineering are also introduced in the first half of this chapter, and fuel design follows. Fuel irradiation behavior, trend of fuel design, and fuel technical standards are introduced as parts of fuel design. The development of cladding for the improvement of fuel burnup and reliability, fuel physical property measurement, and fuel irradiation testing are introduced as fuel development topics in the second half of this chapter.


Fuel Assembly Fuel Pellet Fission Cross Section Coolant Flow Rate Plutonium Content 
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Further Readings

  1. “Plutonium Fuel Technology” (in Japanese), Atomic Energy Society of Japan, (1998)Google Scholar
  2. “Heat-transfer Engineering Data Document”, Fourth Edition (in Japanese), Japan Society of Mechanical Engineers, (1986)Google Scholar
  3. M. Akiyama (ed.), “Nuclear Thermal Engineering” (in Japanese), Nuclear Engineering Series, (University of Tokyo Press. Inc., 1978)Google Scholar
  4. H. Bailly, D. Menessier, C. Prunier (eds.), “The nuclear fuel of pressurized water reactors and fast reactors design and behavior” (in Japanese), K. Konno, Translator, Maruzen Planet Co. LtdGoogle Scholar
  5. N.E. Todreas, M.S. Kazimi, Nuclear systems 1: Thermal hydraulic fundamentals, (Taylor & Francis Inc. 1990)Google Scholar
  6. N.E. Todreas, M.S. Kazimi, N. Todreas, Nuclear systems 2: Elements of thermal hydraulic design, (Taylor & Francis Inc. 1990)Google Scholar
  7. M. Hori (ed.), “Basic Fast Reactor Technology” (in Japanese), Nikkan Kogyo Shimbun, Ltd., (1993)Google Scholar
  8. Edited by Sodium Educational Committee, “Sodium Technology Handbook” (in Japanese), Japan Nuclear Cycle Development Institute, JNC TN9410 2005–011 (2005)Google Scholar
  9. Y. Oka, Writer and Editor, “Textbook of reactor design” (in Japanese), Ohmsha, LtdGoogle Scholar
  10. N. Nakae, et al., “Scope and content covered by LMFBR fuel design” (in Japanese), J. At. Energy Soc. Japan 53(2), 40 (2011)Google Scholar
  11. N. Nakae, et al., “Basic concept of fuel safety design and assessment for sodium-cooled fast reactor”, JNES RE Report Series, JNES-RE-2012-0022, March, 2013Google Scholar
  12. A.E. Waltar, A.B. Reynolds (eds.), Fast Breeder Reactors, (Pergamon Press, Oxford, 1981)Google Scholar
  13. N. Nakae, “Management of plutonium content based on reactivity of each plutonium isotope”, J. Nucl. Sci. Technol. 45(4), 361–366 (2006)Google Scholar


  1. 1.
    Japan Atomic Energy Agency, “Application for reactor establishment license: Prototype fast breeder reactor Monju” (in Japanese)Google Scholar
  2. 2.
    Japan Nuclear Cycle Development Institute, JNC Technical Report, No. 18 (2003) (in Japanese)Google Scholar
  3. 3.
    JAEA Nuclear Data CenterGoogle Scholar
  4. 4.
    T. Takeda et al., Study on detailed calculation and experiment methods of neutronics, fuel materials, and thermal hydraulics for a commercial type Japanese sodium-cooled fast reactor. Sci. Technol. Nuclear Install. 2012 (2012)Google Scholar
  5. 5.
    G. Chiba et al., JENDL4.0 benchmarking for fission reactor applications. J. Nucl. Sci. Technol. 48(2), 172–187 (2011)CrossRefGoogle Scholar
  6. 6.
    F. Yamada et al., Development of natural circulation analytical model in SUPER-COPD code and evaluation of core cooling capability in Monju during a station blackout. Nucl. Technol. 188, 292 (2014)Google Scholar
  7. 7.
    S.J. Zinkle, “Advanced materials for future nuclear plants”, fission energy workshop: Opportunities for fundamental research and breakthrough in fission, Global Climate & Energy Project, MIT, Cambridge, MA, November 29–30 (2007)Google Scholar
  8. 8.
    M. Kato et al., Solidus and liquidus temperatures in the UO2-PuO2 system. J. Nucl. Mater. 373, 237–245 (2008)CrossRefGoogle Scholar
  9. 9.
    K. Morimoto et al., Thermal Conductivity of (U, Pu, Am)O2 Solid Solution. J. Alloys Compd. 452, 54–60 (2008)Google Scholar
  10. 10.
    M. Inoue et al., Power-to-melts of uranium-plutonium oxide fuel pins at beginning-of-life condition in the experimental fast reactor JOYO. J. Nucl. Mater. 323, 108–122 (2003)CrossRefGoogle Scholar

Copyright information

© Springer Nature Singapore Pte Ltd. 2017

Authors and Affiliations

  • Nobuo Nakae
    • 1
    Email author
  • Toshikazu Takeda
    • 2
  • Hiroyuki Ohshima
    • 3
  1. 1.Nuclear Regulation AuthorityTokyoJapan
  2. 2.University of FukuiFukuiJapan
  3. 3.Japan Atomic Energy AgencyIbarakiJapan

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