Skip to main content
SpringerLink
Log in

Development and application of a multi-physics and multi-scale coupling program for lead-cooled fast reactor

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

Abstract

In this study, a multi-physics and multi-scale coupling program, Fluent/KMC-sub/NDK, was developed based on the user-defined functions (UDF) of Fluent, in which the KMC-sub-code is a sub-channel thermal–hydraulic code and the NDK code is a neutron diffusion code. The coupling program framework adopts the "master–slave" mode, in which Fluent is the master program while NDK and KMC-sub are coupled internally and compiled into the dynamic link library (DLL) as slave codes. The domain decomposition method was adopted, in which the reactor core was simulated by NDK and KMC-sub, while the rest of the primary loop was simulated using Fluent. A simulation of the reactor shutdown process of M2LFR-1000 was carried out using the coupling program, and the code-to-code verification was performed with ATHLET, demonstrating a good agreement, with absolute deviation was smaller than 0.2%. The results show an obvious thermal stratification phenomenon during the shutdown process, which occurs 10 s after shutdown, and the change in thermal stratification phenomena is also captured by the coupling program. At the same time, the change in the neutron flux density distribution of the reactor was also obtained.

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

Access this article

Log in via an institution

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

Instant access to the full article PDF.

Institutional subscriptions

Fig. 1
Fig. 2
Fig. 3
Fig. 4
Fig. 5
Fig. 6
Fig. 7
Fig. 8
Fig. 9

Similar content being viewed by others

References

  1. D.L. Aumiller, E.T. Tomlinson, R.C. Bauer, A coupled RELAP5-3D/CFD methodology with a proof-of-principle calculation. Nucl. Eng. Des. 205(1–2), 83–90 (2001). https://doi.org/10.1016/S0029-5493(00)00370-8

    Article  Google Scholar 

  2. D.L. Aumiller, E.T. Tomlinson, W.L. Weaver, An integrated relap5-3d and multiphase cfd code system utilizing a semi-implicit coupling technique. Nucl. Eng. Des. 216(1–3), 77–87 (2002). https://doi.org/10.1016/S0029-5493(01)00522-2

    Article  Google Scholar 

  3. A. Papukchiev, G. Lerchl, C. Waata et al., Extension of the simulation capabilities of the 1D system code ATHLET by coupling with the 3D CFD software package ANSYS CFX. Paper presented at the proceedings of the 13th international topical meeting on nuclear reactor thermal-hydraulics (NURETH-13), Kanazawa City, Japan, September 27 October 2 2009

  4. A. Papukchiev, M. Jeltsov, K. Kööp et al., Comparison of different coupling CFD–STH approaches for pre-test analysis of a TALL-3D experiment. Nucl. Eng. Des. 290, 135–143 (2015). https://doi.org/10.1016/j.nucengdes.2014.11.008

    Article  Google Scholar 

  5. R. Baviere, N. Tauveron, F. Perdu et al., A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO_U. Preliminary validation on the Phénix reactor natural circulation test. Nucl. Eng. Des. 277, 124–137 (2014). https://doi.org/10.1016/j.nucengdes.2014.05.031

    Article  Google Scholar 

  6. Z.R. Zou, C. Shen, X.L. Zhang et al., 3D thermal hydraulic characteristics analysis of pool-type upper plenum for lead-cooled fast reactor with multi-scale coupling program. Nucl. Eng. Des. 370, 110892 (2020)

    Article  Google Scholar 

  7. C. Wang, L.K. Cao, X.L. Zhang et al., Development and application of CFD and subchannel coupling analysis code for lead-cooled fast reactor. Int. J. Energy Res. 43(14), 8447–8462 (2019). https://doi.org/10.1002/er.4845

    Article  Google Scholar 

  8. X.B. Zhang, Z. Qin, H.L. Chen, Development and validation of a coupled neutron diffusion-thermal-hydraulic calculation procedure for fast reactor applications. Ann. Nucl. Energy 139, 107234 (2020). https://doi.org/10.1016/j.anucene.2019.107243

    Article  Google Scholar 

  9. K. Wang, S.C. Liu, Z.G. Li et al., Analysis of BEAVRS two-cycle benchmark using RMC based on full core detailed model. Prog. Nucl. Energy 98, 301–312 (2017). https://doi.org/10.1016/j.pnucene.2017.04.009

    Article  Google Scholar 

  10. S.C. Liu, J.G. Liang, Q. Wu et al., BEAVRS full core burnup calculation in hot full power condition by RMC code. Ann. Nucl. Energy 101, 434–446 (2017). https://doi.org/10.1016/j.pnucene.2017.04.009

    Article  Google Scholar 

  11. C. Fiorina, I. Clifford, M. Aufiero et al., GeN-Foam: a novel OpenFOAM® based multi-physics solver for 2D/3D transient analysis of nuclear reactors. Nucl. Eng. Des. 294, 24–37 (2015). https://doi.org/10.1016/j.nucengdes.2015.05.035

    Article  Google Scholar 

  12. C. Fiorina, N. Kerkar, K. Mikityuk et al., Development and verification of the neutron diffusion solver for the GeN-Foam multi-physics platform. Ann. Nucl. Energy 96, 212–222 (2016). https://doi.org/10.1016/j.anucene.2016.05.023

    Article  Google Scholar 

  13. M.R. Altahhan, S. Bhaskar, D. Ziyad et al., Preliminary design and analysis of liquid fuel molten salt reactor using multi-physics code GeN-Foam. Nucl. Eng. Des. 369, 110826 (2020). https://doi.org/10.1016/j.nucengdes.2020.110826

    Article  Google Scholar 

  14. Z. Chen, X.N. Chen, A. Rineiski et al., Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses. Ann. Nucl. Energy 83, 41–49 (2015). https://doi.org/10.1016/j.anucene.2015.03.023

    Article  Google Scholar 

  15. A.R. Ansys®, Release 15.0, ANSYS Fluent UDF manual (ANSYS Inc, Canonsburg, PA, USA, 2013)

    Google Scholar 

  16. S.Z. Li, L.K. Cao, M.S. Khan et al., Development of a sub-channel thermal-hydraulic analysis code and its application to lead cooled fast reactor. Appl. Therm. Eng. 117, 443–451 (2017). https://doi.org/10.1016/j.applthermaleng.2017.02.044

    Article  Google Scholar 

  17. H.L. Chen, X.L. Zhang, Y.S. Zhao et al., Preliminary design of a medium-power modular lead-cooled fast reactor with the application of optimization methods. Int. J. Energy Res. 42, 3643–3657 (2018). https://doi.org/10.1002/er.4112

    Article  Google Scholar 

  18. C. Shen, X.L. Zhang, C. Wang et al., Transient safety analysis of M2LFR-1000 reactor using ATHLET. Nucl. Eng. Technol. 51, 116–124 (2019). https://doi.org/10.1016/j.net.2018.08.011

    Article  Google Scholar 

  19. X. Luo, X.L. Zhang, S. Wang et al., Automated core design code development for a lead-cooled fast reactor and its core optimization. Int. J. Energy Res. 45, 11721–11734 (2021). https://doi.org/10.1002/er.5553

    Article  Google Scholar 

  20. A. Wielenberg, C. Bals, Nuclear thermal hydraulics with the AC2 system code package. Nucl. Power Plant Des. Anal. Codes 12, 277–311 (2021). https://doi.org/10.1016/B978-0-12-818190-4.00012-7

    Article  Google Scholar 

  21. S.D. Shyam, K.S. Akshaya, G. Padmakumar et al., CFD analysis of thermal stratification and sensitivity study of model parameters for k–ɛ model in a cylindrical hot plenum. Nucl. Eng. Des. 250, 417–435 (2012). https://doi.org/10.1016/j.nucengdes.2012.04.008

    Article  Google Scholar 

  22. S. Ohno, H. Ohshima, H. Ohki et al., Validation of numerical simulation method for thermal stratification in reactor vessel upper plenum of fast reactor. Paper presented at the proceeding of 6th Japan-Korea symposium on nuclear thermal-hydraulics and safety (NTHAS6), Okinawa, Japan, 24–27 Nov 2008

  23. M. Vanderhaegen, J. Vierendeels, B. Arien, CFD analysis of the MYRRHA primary cooling system. Nucl. Eng. Des. 241(3), 775–784 (2011). https://doi.org/10.1016/j.nucengdes.2010.12.009

    Article  Google Scholar 

  24. F. Roelofs, Thermal hydraulics aspects of liquid metal cooled nuclear reactors (M. Woodhead Publishing, USA, 2018)

    Google Scholar 

  25. B. Cai, H.F. Gu, Y. Weng et al., Numerical investigation on the thermal stratification in a pressurizer surge line. Ann. Nucl. Energy 101, 293–300 (2017). https://doi.org/10.1016/j.anucene.2016.11.024

    Article  Google Scholar 

Download references

Acknowledgements

We thank GRS for providing the ATHLET code.

Author information

Authors and Affiliations

  1. School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, 230026, China

    Xiao Luo, Chi Wang, Ze-Ren Zou, Lian-Kai Cao, Shuai Wang & Hong-Li Chen

  2. China Nuclear Power Technology Research Institute Co. Ltd, Shenzhen, 518000, China

    Zhao Chen

Authors
  1. Xiao Luo
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. Chi Wang
    View author publications

    You can also search for this author in PubMed Google Scholar

  3. Ze-Ren Zou
    View author publications

    You can also search for this author in PubMed Google Scholar

  4. Lian-Kai Cao
    View author publications

    You can also search for this author in PubMed Google Scholar

  5. Shuai Wang
    View author publications

    You can also search for this author in PubMed Google Scholar

  6. Zhao Chen
    View author publications

    You can also search for this author in PubMed Google Scholar

  7. Hong-Li Chen
    View author publications

    You can also search for this author in PubMed Google Scholar

Contributions

All authors contributed to the study conception and design. Material preparation, data collection and analysis were performed by Xiao Luo, Chi Wang, Ze-Ren Zou, Lian-Kai Cao, Shuai Wang, Zhao Chen, and Hong-Li Chen. The first draft of the manuscript was written by Xiao Luo and all authors commented on previous versions of the manuscript. All authors read and approved the final manuscript.

Corresponding authors

Correspondence to Zhao Chen or Hong-Li Chen.

Additional information

This work was supported by Science and Technology on Reactor System Design Technology Laboratory, Chengdu, China (LRSDT2020106).

Rights and permissions

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Luo, X., Wang, C., Zou, ZR. et al. Development and application of a multi-physics and multi-scale coupling program for lead-cooled fast reactor. NUCL SCI TECH 33, 18 (2022). https://doi.org/10.1007/s41365-022-01008-y

Download citation

  • Received:

  • Revised:

  • Accepted:

  • Published:

  • DOI: https://doi.org/10.1007/s41365-022-01008-y

Keywords

Search

Navigation