Skip to main content
Log in

CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam

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

Abstract

Steam generator tube rupture (SGTR) accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors. When the steam generator tube breaks close to the main pump, water vapor will enter the reactor core, resulting in a two-phase flow of heavy liquid metal and water vapor in fuel assemblies. The thermal-hydraulic problems caused by the SGTR accident may seriously threaten reactor core’s safety performance. In this paper, the open-source CFD calculation software OpenFOAM was used to encapsulate the improved Euler method into the self-developed solver LBEsteamEulerFoam. By changing different heating boundary conditions and inlet coolant types, the two-phase flow in the fuel assembly with different inlet gas content was simulated under various accident conditions. The calculation results show that the water vapor may accumulate in edge and corner channels. With the increase in inlet water vapor content, outlet coolant velocity increases gradually. When the inlet water vapor content is more than 15%, the outlet coolant temperature rises sharply with strong temperature fluctuation. When the inlet water vapor content is in the range of 5–20%, the upper part of the fuel assembly will gradually accumulate to form large bubbles. Compared with the VOF method, Euler method has higher computational efficiency. However, Euler method may cause an underestimation of the void fraction, so it still needs to be calibrated with future experimental data of the two-phase flow in fuel assembly.

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

Similar content being viewed by others

Data availability

The data that support the findings of this study are openly available in Science Data Bank at https://www.doi.org/10.57760/sciencedb.j00186.00220 and https://cstr.cn/31253.11.sciencedb.j00186.00220.

References

  1. J.A. Lake, The fourth generation of nuclear power. Prog. Nucl. Energy 40, 301–307 (2002). https://doi.org/10.1016/S0149-1970(02)00023-9

    Article  Google Scholar 

  2. J. Wang, W.X. Tian, Y.H. Tian et al., Thermal-hydraulic primary numerical analysis for Pb-Bi Fast Reactor Sub-channel. At. Energy Sci. Technol. 47, 38–42 (2013). (in Chinese)

    Google Scholar 

  3. A. Alemberti, V. Smirnov, C.F. Smith et al., Overview of lead-cooled fast reactor activities. Prog. Nucl. Energy 77, 300–307 (2014). https://doi.org/10.1016/j.pnucene.2013.11.011

    Article  Google Scholar 

  4. L. Zhang, Y.W. Yang, Y.C. Cao, Preliminary physics study of the Lead-Bismuth-Eutectic spallation target for China Initiative Accelerator Driven Subcritical System. Nucl. Sci. Tech. 27, 120 (2016). https://doi.org/10.1007/s41365-016-0114-6

    Article  Google Scholar 

  5. Z.Q. Liu, Z.L. Zhao, Y.W. Yang et al., Shielding calculation of LBE target flow pipeline in ADS. Nuclear Techniques 41, 030604 (2018). https://doi.org/10.11889/j.0253-3219.2018.hjs.41.030604 (in Chinese)

    Article  Google Scholar 

  6. J. Wen, T.J. Peng, X.K. Fan et al., Analysis and optimization of flow distribution for the reactor core of China initiative accelerator driven system. Nuclear Techniques 43, 070601 (2020). https://doi.org/10.11889/j.0253-3219.2020.hjs.43.070601 (in Chinese)

    Article  Google Scholar 

  7. L. Cinotti, C.F. Smith, H. Sekimoto et al., Lead-cooled system design and challenges in the frame of generation lV international forum. J. Nucl. Mater. 415, 245–253 (2011). https://doi.org/10.1016/j.jnucmat.2011.04.042

    Article  ADS  Google Scholar 

  8. J. Xue, Rupture Failure Analysis and Corrosion Mechanism Research of Steam Generator Tubes. Thesis M. A. Harbin Engineering University (2007) (in Chinese)

  9. T. N. Dinh, Multiphase flow phenomena of steam generator tube rupture in a Lead-cooled reactor system: a scooping analysis, in Paper Presented at International Congress on Advances in Nuclear Power Plants (Nice, 2007)

  10. Y. Sibamoto, Y. Kukita, H. Nakamura, Small-scale experiment on subcooled water jet injection into molten alloy by using fluid temperature-phase coupled measurement and visualization. J. Nucl. Sci. Technol. 44, 1059–1069 (2007). https://doi.org/10.1080/18811248.2007.9711347

    Article  ADS  Google Scholar 

  11. R. Sa, M. Takahashi, K. Moriyama, Study on fragmentation behavior of liquid lead alloy droplet in water. Prog. Nucl. Energy 53, 895–901 (2011). https://doi.org/10.1016/j.pnucene.2011.05.003

    Article  Google Scholar 

  12. R. Sa, M. Takahashi, Experimental study on thermal interaction of ethanol jets in high temperature fluorinert. J. Power Energy Syst. 6, 314–323 (2012). https://doi.org/10.1299/jpes.6.314

    Article  ADS  Google Scholar 

  13. V. Dostal, M. Takahashi, Boiling heat transfer behavior of lead-bismuth-steam-water direct contact two-phase flow. Prog. Nucl. Energy 50, 625–630 (2008). https://doi.org/10.1016/j.pnucene.2007.11.058

    Article  Google Scholar 

  14. W.L. Huang, R.Y. Sa, D.N. Zhou et al., Experimental study on fragmentation behaviors of molten LBE and water contact interface. Nucl. Sci. Tech. 26, 060601 (2015). https://doi.org/10.13538/j.1001-8042/nst.26.060601

    Article  Google Scholar 

  15. W. Huang, D. Zhou, R. Sa et al., Experimental study on thermal-hydraulic behaviour of LBE and water interface. Prog. Nucl. Energy. 99, 1–10 (2017). https://doi.org/10.1016/j.pnucene.2017.04.005

    Article  Google Scholar 

  16. Z.J. Deng, S.B. Cheng, H. Cheng, Experimental investigation on pressure-buildup characteristics of a water lump immerged in a molten lead pool. Nucl. Sci. Tech. 34, 35 (2023). https://doi.org/10.1007/s41365-023-01188-1

    Article  Google Scholar 

  17. R. Sa, Study on thermal-hydraulic behaviors in direct contact of high temperature lead alloy and subcooled water. Thesis M. A. Tokyo Institute of Technology (2012)

  18. S. Wang, M. Flad, W. Maschek et al., Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling. Prog. Nucl. Energy 50, 363–369 (2008). https://doi.org/10.1016/j.pnucene.2007.11.018

    Article  Google Scholar 

  19. A. Ciampichetti, A. Nevo, G. Bandini et al., SG tube rupture in LFR, in Paper Presented at the International Workshop on Innovative Nuclear Reactors Cooled by Heavy Liquid Metals: Status and Perspectives (Pisa, 2012)

  20. Z.X. Gu, G. Wang, Y.Q. Bai et al., Preliminary investigation on the primary heat exchanger lower head rupture accident of forced circulation LBE-cooled fast reactor. Ann. Nucl. Energy 81, 84–90 (2015). https://doi.org/10.1016/j.anucene.2015.03.018

    Article  Google Scholar 

  21. M. Jeltsov, W. Villanueva, P. Kudinov, Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core. Nucl. Eng. Des. 328, 255–265 (2018). https://doi.org/10.1016/j.nucengdes.2018.01.006

    Article  Google Scholar 

  22. Z.X. Gu, Q.X. Zhang, Y. Gu et al., Verification of a self-developed CFD-based multi-physics coupled code MPC-LBE for LBE-cooled reactor. Nucl. Sci. Tech. 32, 52 (2021). https://doi.org/10.1007/s41365-021-00887-x

    Article  Google Scholar 

  23. S.Y. Liu, D.L. Yu, H.P. Mei et al., Turbulent-Prandtl-number models for liquid lead-bismuth in triangular rod bundles. Nuclear Techniques 45, 030604 (2022). https://doi.org/10.11889/j.0253-3219.2022.hjs.45.030604. (in Chinese)

    Article  Google Scholar 

  24. Y.X. Li, L. Meng, Z.N. Huang et al., Study on the effects from spacer wires on coolant flow within a CiADS fuel assembly. Ann. Nucl. Energy 183, 109647 (2023). https://doi.org/10.1016/j.anucene.2022.109647

    Article  Google Scholar 

  25. T.T. Zhou, S.Y. Liu, J. Yu, Friction pressure drop model for wire-wrapped rod bundles in full flow. Nuclear Techniques 46, 060604 (2023). https://doi.org/10.11889/j.0253-3219.2023.hjs.46.060604. (in Chinese)

    Article  Google Scholar 

  26. T. Suzuki, Y. Tobita, S. Kondo et al., Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III. Nucl. Eng. Des. 220, 207–223 (2003). https://doi.org/10.1016/S0029-5493(02)00349-7

    Article  Google Scholar 

  27. T. Wang, P.C. Zhao, Z.J. Liu et al., Thermal-hydraulic analysis method for the annular fuel structure of lead-bismuth reactor. Nuclear Techniques 44, 100604 (2021). https://doi.org/10.11889/j.0253-3219.2021.hjs.44.100604. (in Chinese)

    Article  Google Scholar 

  28. L.T. Shen, X. Chai, X. Cheng, Numerical simulation of gas injected bubble dynamics from single submerged orifice. Nucl. Power Eng. 41, 194–197 (2020). (in Chinese)

    Google Scholar 

  29. O. Shoham, Flow pattern transition and characterization in gas-liquid two phase flow in inclined pipes. Thesis M. A. Tel Aviv University (1982)

  30. M. Pourtousi, J.N. Sahu, P. Ganesan, Effect of interfacial forces and turbulence models on predicting flow pattern in-side the bubble column. Chem. Eng. Process. 75, 38–47 (2014). https://doi.org/10.1016/j.cep.2013.11.001

    Article  Google Scholar 

  31. R. Adoua, D. Legendre, J. Magnaudet, Reversal of the lift force on an oblate bubble in a weakly viscous linear shear flow. J. Fluid Mech. 628, 23–41 (2009). https://doi.org/10.1017/S0022112009006090

    Article  ADS  MATH  Google Scholar 

  32. S.P. Antal, R.T. Lahey Jr., J.E. Flaherty, Analysis of phase distribution in fully developed laminar bubbly two-phase flow. Int. J. Multiphase Flow 17, 635–652 (1991). https://doi.org/10.1016/0301-9322(91)90029-3

    Article  MATH  Google Scholar 

  33. M.A.L. De Bertodano, Two fluid model for two-phase turbulent jets. Nucl. Eng. Des. 179, 65–74 (1998). https://doi.org/10.1016/S0029-5493(97)00244-6

    Article  Google Scholar 

  34. D. Zhang, N.G. Deen, J.A.M. Kuipers, Numerical simulation of the dynamic flow behavior in a bubble column: a study of closures for turbulence and interface forces. Chem. Eng. Sci. 61, 7593–7608 (2006). https://doi.org/10.1016/j.ces.2006.08.053

    Article  Google Scholar 

  35. X.K. Su, Numerical study on turbulent heat transfer of liquid lead bismuth based on an isotropic four-equation model. Thesis M. A. University of Chinese Academy of Sciences (Institute of Modern Physics, Chinese Academy of Sciences) (2022) (in Chinese)

  36. X. Cheng, N.I. Tak, Investigation on turbulent heat transfer to lead-bismuth eutectic flows in circular tubes for nuclear applications. Nucl. Eng. Des. 236, 385–393 (2006). https://doi.org/10.1016/j.nucengdes.2005.09.006

    Article  Google Scholar 

  37. J.T. Liu, Research on subchannel analysis method of lead-based fast reactor fuel assembly with wire spacers for CiADS. Thesis M. A. University of Chinese Academy of Sciences (Institute of Modern Physics, Chinese Academy of Sciences) (2021) (in Chinese)

  38. T.J. Peng, L. Gu, D.W. Wang et al., Conceptual design of subcritical reactor for China initiative accelerator driven system. At. Energy Sci. Technol. 51, 2235–2241 (2017). (in Chinese)

    Google Scholar 

  39. Z.F. Ge, T. Zhou, Y.Q. Bai et al., Thermal-hydraulic analysis in wire-wrapped fuel assembly for china lead-based research reactor. At. Energy Sci. Technol. 49, 167–173 (2015). (in Chinese)

    Google Scholar 

  40. OECD, Handbook on Lead-Bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal Hydraulics and Technologies (OECD, Paris, 2007)

    Google Scholar 

  41. W. Wagner, J.R. Cooper, The IAPWS industrial formulation 1997 for the thermodynamic properties of water and steam. Eng. Gas. Turb. Power 122, 150–182 (2000). https://doi.org/10.1007/978-3-540-74234-0_3

    Article  Google Scholar 

  42. J. Pacio, M. Daubner, F. Fellmoser et al., Experimental study of heavy-liquid metal (LBE) flow and heat transfer along a hexagonal 19-rod bundle with wire spacers. Nucl. Eng. Des. 301, 111–127 (2016). https://doi.org/10.1016/j.nucengdes.2016.03.003

    Article  Google Scholar 

  43. J. Deng, Q. Lu, D. Wu et al., Sub-channel code development of lead-bismuth eutectic fast reactor available for multiple fuel assembly structures. Ann. Nucl. Energy 149, 107769 (2020). https://doi.org/10.1016/j.anucene.2020.107769

    Article  Google Scholar 

  44. S. Yang, Y.P. Zhang, Study on two-phase flow in fluid channel of lead-based fast reactor based on OpenFOAM. At. Energy Sci. Technol. 54, 1582–1588 (2020). (in Chinese)

    Google Scholar 

Download references

Author information

Authors and Affiliations

Authors

Contributions

All authors contributed to the study conception and design. Material preparation, data collection, and analysis were performed by Yun-Xiang Li, Lu Meng, Zi-Nan Huang, Song Li, Di-Si Wang, Bo Liu, You-Peng Zhang, Tian-Ji Peng, Lu Zhang, Xing-Kang Su, and Wei Jiang. The first draft of the manuscript was written by Yun-Xiang Li, and all authors commented on previous versions of the manuscript. All authors read and approved the final manuscript.

Corresponding author

Correspondence to You-Peng Zhang.

Ethics declarations

Conflict of interest

The authors declare that they have no competing interests.

Additional information

This work was supported partly by the Ministry of Science and Technology of the People's Republic of China (No. 2020YFB1902100) and the Shanghai Municipal Commission of Economy and Informatization (No. GYQJ-2018-2-02).

Rights and permissions

Springer Nature or its licensor (e.g. a society or other partner) holds exclusive rights to this article under a publishing agreement with the author(s) or other rightsholder(s); author self-archiving of the accepted manuscript version of this article is solely governed by the terms of such publishing agreement and applicable law.

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Li, YX., Meng, L., Li, S. et al. CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam. NUCL SCI TECH 34, 157 (2023). https://doi.org/10.1007/s41365-023-01312-1

Download citation

  • Received:

  • Revised:

  • Accepted:

  • Published:

  • DOI: https://doi.org/10.1007/s41365-023-01312-1

Keywords

Navigation