Skip to main content

Advertisement

Log in

In-core fuel management strategy for the basket-fuel-assembly molten salt reactor

  • Published:
Nuclear Science and Techniques Aims and scope Submit manuscript

Abstract

Molten salt reactor, with good economics and inherent reliability, is one of the six types of Generation IV candidate reactors. The Basket-Fuel-Assembly Molten Salt Reactor (BFAMSR) is a new concept design based on fuel assemblies composed of fuel pebbles made of TRISO-coated particles. Four refueling patterns, similar to the fuel management strategy for water reactors, are designed and analyzed for BFAMSR in terms of economy and security. The MCNPX is employed to calculate the parameters, such as the total duration time, cycle length, discharge burnup, total discharge quantity of 235U, total discharge quantity of 239Pu, neutron flux distribution and power distribution. The in–out loading pattern has the highest burnup and duration time, the worst neutron flux and power distribution and the lowest neutron leakage. The out–in pattern possesses the most uniform neutron flux distribution, the lowest burnup and total duration time, and the highest neutron leakage. The out–in partition alternate pattern has slightly higher burnup, longer total duration time and smaller neutron leakage than that of the out–in loading pattern at the cost of sacrificing some neutron flux distribution and power distribution. However, its alternative distribution of fuel elements cut down the refueling time. The low-leakage pattern is the second highest in burnup, and total duration time, and its neutron flux and power distributions are the second most uniform.

This is a preview of subscription content, log in via an institution to check access.

Access this article

Price excludes VAT (USA)
Tax calculation will be finalised during checkout.

Instant access to the full article PDF.

Fig. 1
Fig. 2
Fig. 3
Fig. 4
Fig. 5
Fig. 6
Fig. 7
Fig. 8
Fig. 9
Fig. 10
Fig. 11
Fig. 12
Fig. 13

Similar content being viewed by others

References

  1. DOE, A technology roadmap for Generation IV nuclear energy systems. Nuclear Energy Research Advisory Committee and the Generation IV International Forum, GIF-002-00

  2. K. Nagy, D. Lathouwers, C.G.A. T’Joen et al., Steady-state and dynamic behavior of a moderated molten salt reactor. Ann. Nucl. Energy 64, 365–379 (2014). doi:10.1016/j.anucene.2013.08.009

    Article  Google Scholar 

  3. H.G. MacPherson, The molten salt reactor adventure. Nucl. Sci. Eng. 90, 374–380 (1985). doi:10.13182/NSE90-374

    Article  Google Scholar 

  4. J. Serp, M. Allibert, O. Benes et al., The molten salt reactor (MSR) in generation IV: overview and perspectives. Prog. Nucl. Energy 77, 308–319 (2014). doi:10.1016/j.pnucene.2014.02.014

    Article  Google Scholar 

  5. E.S. Bettis, R.W. Schroeder, G.A. Cristy et al., The aircraft reactor experiment design and construction. Nucl. Sci. Eng. 2(6), 804–825 (1957)

    Article  Google Scholar 

  6. M.W. Rosenthal, P.R. Kasten, R.B. Briggs, Molten-salt reactors-history, status, and potential. Nucl. Appl. Technol. 8, 107–117 (1970)

    Article  Google Scholar 

  7. P. Avigni, B. Petrovic, Fuel element and full core thermal–hydraulic analysis of the AHTR for the evaluation of the LOFC transient. Ann. Nucl. Energy 64, 499–510 (2014). doi:10.1016/j.anucene.2013.05.029

    Article  Google Scholar 

  8. V.K. Varma, D.E. Holcomb, F.J. Peretz et al., AHTR mechanical, structural, and neutronic preconceptual design. ORNL/TM-2012/320, ORNL. doi:10.2172/1054145

  9. M. Fratoni, E. Greenspan, Neutronic feasibility assessment of liquid salt–cooled pebble bed reactors. Nucl. Sci. Eng. 168(1), 1–22 (2011). doi:10.13182/NSE10-38

    Article  Google Scholar 

  10. M.H. Jiang, H.J. Xu, Z.M. Dai, Advanced fission energy program-TMSR nuclear energy system. Bull. Chin. Acad. Sci. 27(3), 366–374 (2012). doi:10.3969/j.issn.1000-3045.2012.03.016. (in Chinese)

    Google Scholar 

  11. C. Xue, H.Q. Zhang, Z.Y. Zhu et al., Design of fuel assembly for molten-salt-cooled reactors. Nucl. Tec. 39(090602–1), 090602–090608 (2016). doi:10.11889/j.0253-3219.2016.hjs.39.090602. (in Chinese)

    Google Scholar 

  12. Z.S. Xie, Nuclear Reactor Physics Analysis (Xi’an Jiaotong University Press, Xi’An, 2004), pp.253–285. (in Chinese)

  13. V.N. Bukanov, V.L. Demekhin, A.A. Korennoi, Use of low neutron leakage fuel loads to decrease the radiation load on a VVÉR-1000 vessel. At. Energy. 101(2), 544–548 (2006). doi:10.1007/s10512-006-0128-y

    Article  Google Scholar 

  14. A. Talamo, W. Gudowski, Adapting the deep burn in-core fuel management strategy for the gas turbine—modular helium reactor to a uranium–thorium fuel. Ann. Nucl. Energy 32, 1750–1781 (2005). doi:10.1016/j.anucene.2005.07.002

    Article  Google Scholar 

  15. A. Thakur, B. Singh, P.D. Krishnani, In-core fuel management for AHWR. Ann. Nucl. Energy 57, 47–58 (2013). doi:10.1016/j.anucene.2013.01.034

    Article  Google Scholar 

  16. N.J. Hill, G.T. Parks, Pressurized water reactor in-core nuclear fuel management by tabu search. Ann. Nucl. Energy 75, 64–71 (2015). doi:10.1016/j.anucene.2014.07.051

    Article  Google Scholar 

  17. T.J. Downar, A. Sesonske, Light water reactor fuel cycle optimization: theory versus practice. Adv. Nucl. Sci. Technol. 20(20), 71–126 (1988). doi:10.1007/978-1-4613-9925-4_2

    Google Scholar 

  18. J.S. Hendricks, G.W. Mckinney, M.L. Fensin, MCNPX 2.6.0 Extensions. LA-UR-08-2216, Los Alamos National Laboratory (2008)

  19. K. Shibata, O. Iwamoto, T. Nakagawa et al., JENDL-4.0: a new library for nuclear science and engineering. J. Nucl. Sci. Technol. 48, 1–30 (2011). doi:10.1080/18811248.2011.9711675

    Article  Google Scholar 

  20. Z.G. Ge, Z.X. Zhao, H.H. Xia et al., The updated version of Chinese evaluated nuclear data library (CENDL-3.1). J. Korean Phys. Soc. 59, 1052–1056 (2011). doi:10.3938/jkps.59.1052

    Article  Google Scholar 

  21. K. Allen, T. Knight, S. Bays, Benchmark of advanced burner test reactor model using MCNPX 2.6.0 and ERANOS 2.1. Prog. Nucl. Energy 53, 633–644 (2011). doi:10.1016/j.pnucene.2011.01.007

    Article  Google Scholar 

Download references

Author information

Authors and Affiliations

Authors

Corresponding author

Correspondence to Zhi-Yong Zhu.

Additional information

This work was supported by the Strategic Priority Program of Chinese Academy of Sciences (No.XDA02030200).

Rights and permissions

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Xue, C., Zhu, ZY., Zhang, HQ. et al. In-core fuel management strategy for the basket-fuel-assembly molten salt reactor. NUCL SCI TECH 28, 130 (2017). https://doi.org/10.1007/s41365-017-0286-8

Download citation

  • Received:

  • Revised:

  • Accepted:

  • Published:

  • DOI: https://doi.org/10.1007/s41365-017-0286-8

Keywords

Navigation