Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

  • Yu Peng
  • Gui-Feng Zhu
  • Yang ZouEmail author
  • Si-Jia Liu
  • Hong-Jie XuEmail author


Fast reactors based on thorium fuel have enhanced inherent safety. Fluoride salt performs well as a coolant in high-temperature nuclear systems. In this paper, we present a reference core for a large fluoride-salt-cooled solid-fuel fast reactor (LSFR) using thorium–uranium fuel cycle. Neutronics physics of the LSFR reference core is investigated with 2D and 3D in-core fuel management strategy. The design parameters analyzed include the fuel volume fraction, power density level and continuous removal of fission products with 3D fuel shuffling that obtains better equilibrium core performance than 2D shuffling. A self-sustained core is achieved for all cases, and the core of 60% fuel volume fraction at 50 MW/m3 power density is of the best breeding performance (average breeding ratio 1.134). The LSFR core based on thorium fuel is advantageous in its high discharge burn-up of 20–30% fissions per initial heavy metal atom, small reactivity swing over the whole lifetime (to simplify the reactivity control system), the negative reactivity temperature coefficient (intrinsically safe for all cases) and accepted cladding peak radiation damage. The LSFR reactor is a good alternative option for the deployment of a self-sustained thorium-based nuclear system.


Fluoride salts Thorium cycle Fast reactor Core characteristics Equilibrium 


  1. 1.
    A.E. Waltar, D.R. Todd, P.V. Tsvetkov (eds.), Fast Spectrum Reactors (Springer, New York, 2012)Google Scholar
  2. 2.
    A. Demirbas, Options and trends of thorium fuel utilization. Energy Source 27, 597–603 (2005). doi: 10.1080/00908310490448596 CrossRefGoogle Scholar
  3. 3.
    K. Nagy, J.L. Kloosterman, D. Lathouwers et al., Van der Hagen, Parametric studies on the fuel salt composition in thermal molten salt breeder reactors, in The International Conference on Physics of Reactors 2008 (PHYSOR 08), Interlaken, Switzerland, 14–19 September 2008Google Scholar
  4. 4.
    C. Fiorina, J. Krepel, A. Cammi et al., Analysis of thorium and uranium fuel cycles in an iso-breeder lead fast reactor using extended-EQL3D procedure. Ann. Nucl. Energy 53, 492–506 (2013). doi: 10.1016/j.anucene.2012.09.004 CrossRefGoogle Scholar
  5. 5.
    G.C. Li, Y. Zou, C.G. Yu et al., Influences of 7Li enrichment on Th–U fuel breeding for an Improved Molten Salt Fast Reactor (IMSFR). Nucl. Sci. Tech. 28, 97 (2017). doi: 10.1007/s41365-017-0250-7 CrossRefGoogle Scholar
  6. 6.
    C.W. Forsberg, Use of liquid salt coolants to improve fast-reactor economics. Paper presented at the economic analysis of fast reactors American nuclear society annual meeting, Boston, Massachusetts, 24–28 June 2007Google Scholar
  7. 7.
    D.F. Williams, L.M. Toth, K.T. Clarno, Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR), 1st edn. Technical Report ORNL/TM-2006/12 (United States, Department of Energy, 2006), pp. 1–69Google Scholar
  8. 8.
    A. Acır, Neutronic analysis of the laser inertial confinement fusion-fission energy (LIFE) engine using various thorium molten salts. J. Fusion Energy 32, 634–641 (2013). doi: 10.1007/s10894-013-9628-7 CrossRefGoogle Scholar
  9. 9.
    A. Acır, Improvement of the neutronic performance of the PACER fusion concept using thorium molten salt with reactor grade plutonium. J. Fusion Energy 32, 11–14 (2013). doi: 10.1007/s10894-012-9518-4 CrossRefGoogle Scholar
  10. 10.
    M. Übeyli, A. Acır, Neutronic investigation on the ARIES-ST fusion reactor with fissionable molten salts. Energy Convers. Manag. 51, 2531–2534 (2010). doi: 10.1016/j.enconman.2010.05.018 CrossRefGoogle Scholar
  11. 11.
    J.X. Zuo, J.P. Jing, J.S. Bi et al., Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment. Nucl. Sci. Tech. 26, 050602 (2015). doi: 10.13538/j.1001-8042/nst.26.050602 Google Scholar
  12. 12.
    C. Xue, Z.Y. Zhu, H.Q. Zhang et al., In-core fuel management strategy for the basket-fuel-assembly molten salt reactor. Nucl. Sci. Tech. 28, 130 (2017). doi: 10.1007/s41365-017-0286-8 CrossRefGoogle Scholar
  13. 13.
    J. Ruan, B. Xu, M.H. Li et al., A specialized code for operation transient analysis and its application in fluoride salt-cooled high-temperature reactors. Nucl. Sci. Tech. 28, 119 (2017). doi: 10.1007/s41365-017-0268-x CrossRefGoogle Scholar
  14. 14.
    K. Yang, W. Qin, J.G. Chen et al., Neutron excess method for performance assessment of thorium-based fuel in a breed-and-burn reactor with various coolants. Nucl. Sci. Tech. 27, 99 (2016). doi: 10.1007/s41365-016-0096-4 CrossRefGoogle Scholar
  15. 15.
    Z.B. Liu, Y. Liu, G.M. Liu et al., Reactor protection system testing for the solid fuel thorium molten salt reactor. Nucl. Sci. Tech. 27, 123 (2016). doi: 10.1007/s41365-016-0091-9 CrossRefGoogle Scholar
  16. 16.
    G.M. Sun, M.S. Cheng, Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor. Nucl. Sci. Tech. 27, 65 (2016). doi: 10.1007/s41365-016-0070-1 CrossRefGoogle Scholar
  17. 17.
    C. Behar, Technology roadmap update for Generation IV nuclear energy systems. (IOP Publishing Physics Web, 2014), Accessed March 2014
  18. 18.
    G. Locatelli, M. Mancini, N. Todeschini, Generation IV nuclear reactors: current status and future prospects. Energy Policy 61, 1503–1520 (2013). doi: 10.1016/j.enpol.2013.06.101 CrossRefGoogle Scholar
  19. 19.
    C.W. Forsberg, P.F. Peterson, D.F. Williams, Practical aspects of liquid-salt-cooled fast-neutron reactors, in The International Congress on Advances in Nuclear Power Plants 2005: ICAPP 05, Seoul, Korea, 15–19 May 2005Google Scholar
  20. 20.
    C.W. Forsberg, P.F. Peterson, P.S. Pickard, Molten-salt-cooled advanced high-temperature reactor for production of hydrogen and electricity. Nucl. Technol. 144, 289–302 (2003). doi: 10.13182/NT03-1 CrossRefGoogle Scholar
  21. 21.
    C.W. Forsberg, C.L. Brun, D.F. E.M. Lucotte et al., Practical aspects of liquid-salt-cooled fast-neutron reactors, in The International Congress on Advances in Nuclear Power Plants 2007: ICAPP 07, Nice, France, 13–18 May 2007Google Scholar
  22. 22.
    Z. Perkó, S. Pelloni, K. Mikityuk et al., Core neutronics characterization of the GFR2400 gas cooled fast reactor. Prog. Nucl. Energy 83, 460–481 (2015). doi: 10.1016/j.pnucene.2014.09.016 CrossRefGoogle Scholar
  23. 23.
    J. Hou, S. Qvist, E. Greenspan, 3-D fuel shuffling for reduced peak burnup and increased uranium utilization of breed-and-burn reactors, in The International Congress on Advances in Nuclear Power Plants 2015: ICAPP 2015, Nice, France, 03–06 May 2015Google Scholar
  24. 24.
    F. Heidet, E. Greenspan, Neutron balance analysis for sustainability of breed-and-burn reactors. Nucl. Sci. Eng. 171, 13–31 (2012). doi: 10.13182/NSE10-114 CrossRefGoogle Scholar
  25. 25.
    G.F. Zhu, Dissertation, University of Chinese Academy of Sciences, 2015Google Scholar
  26. 26.
    Y. Peng, G.F. Zhu, Y. Zou et al., in Progress in Nuclear Energy, ed. by N. Vishwaraj (Elsevier, Amsterdam, 2017). (in press) Google Scholar
  27. 27.
    R.C. Hoyt, B.W. Rhee, Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes. Technical Report ESG-DOE—13277 ON: DE92004038 (1979). doi: 10.2172/10108445
  28. 28.
    S. Delpech, E.M. Lucotte, D. Heuer et al., Reactor physic and reprocessing scheme for innovative molten salt reactor system. J. Fluor. Chem. 130, 11–17 (2009). doi: 10.1016/j.jfluchem.2008.07.009 CrossRefGoogle Scholar

Copyright information

© Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Chinese Nuclear Society, Science Press China and Springer Nature Singapore Pte Ltd. 2017

Authors and Affiliations

  1. 1.Shanghai Institute of Applied PhysicsChinese Academy of SciencesShanghaiChina
  2. 2.University of Chinese Academy of SciencesBeijingChina

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