Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Applicability research of round tube CHF mechanistic model in rod bundle channel

  • Liu, Wei (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology) ;
  • Peng, Shinian (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology) ;
  • Shan, Jianqiang (Xi'an Jiaotong University, School of Nuclear Science and Technology) ;
  • Jiang, Guangming (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology) ;
  • Liu, Yu (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology) ;
  • Deng, Jian (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology) ;
  • Hu, Ying (Nuclear Power Institute of China, State Key Laboratory of Reactor System Design Technology)
  • Received : 2020.05.05
  • Accepted : 2020.07.15
  • Published : 2021.02.25
Download PDF
⟨ Previous Next ⟩

Abstract

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

Keywords

Acknowledgement

The authors express their appreciation to Nuclear Power Institute of China for their financial support.

References

  1. D. Groeneveld, The critical heat flux story, in: May 12-17: The 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 2013. Pisa, Italy.
  2. Wei Liu, Hideki Nariai, Ultrahigh CHF prediction for subcooled flow boiling based on homogenous nucleation mechanism, J. Heat Tran. 127 (2) (2005) 149-158. https://doi.org/10.1115/1.1844536
  3. L.S. Tong, Heat transfer in water cooled nuclear reactors, Nucl. Eng. Des. 6 (4) (1967) 301-324. https://doi.org/10.1016/0029-5493(67)90111-2
  4. P.B. Whalley, The calculation of dryout in a rod bundle, Int. J. Multiphas. Flow 3 (6) (1977) 501-515. https://doi.org/10.1016/0301-9322(77)90026-X
  5. J. Saito, E.D. Hughes, M.W. Carbon, Multi-Fluid modeling of annular two-phase flow, Nucl. Eng. Des. 50 (2) (1978) 225-271. https://doi.org/10.1016/0029-5493(78)90041-9
  6. T. Mitsutake, H. Terasaka, K. Yoshimura, et al., Subchannel analysis of a critical power test using simulated BWR 8×8 fuel assembly, Nucl. Eng. Des. 122 (1-3) (1990) 235-254. https://doi.org/10.1016/0029-5493(90)90209-G
  7. H. Zhang, G. Hewitt, New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles, Nucl. Eng. Des. 305 (2016) 73-80. https://doi.org/10.1016/j.nucengdes.2016.04.044
  8. Wei Liu, Jianqiang Shan, Shinian Peng, et al., The study of critical heat flux in upflow boiling vertical round tube under high pressure, Sci. Technol. Nucl. Install. (2019) ID3695685.
  9. M. Bruder, G. Bloch, T. Sattelmayer, Critical heat flux in flow boiling-review of the current understanding and experimental approaches, Heat Tran. Eng. 38 (3) (2017) 347-360. https://doi.org/10.1080/01457632.2016.1189274
  10. T.H. Chun, W.P. Baek, S.H. Chang, An integral equation model for critical heat flux at subcooled and low quality flow boiling, Nucl. Eng. Des. 199 (1-2) (2000) 13-29. https://doi.org/10.1016/S0029-5493(00)00224-7
  11. Jianqiang Shan, Bo Zhang, Changying Li, et al., SCWR subchannel code ATHAS development and CANDU-SCWR analysis, Nucl. Eng. Des. 239 (2009) 1979-1987. https://doi.org/10.1016/j.nucengdes.2009.05.008
  12. Wei Liu, B.A.I. Ning, Yuan-bing Zhu, et al., Research of reactor thermal-hydraulics sub-channel analysis code ATHAS, Nucl. Sci. Eng. 34 (1) (2014) 59-66.
  13. Jun Huang, Junli Gou, Wenjie Ding, et al., Development of a sub-channel analysis code based on the two-fluid model and its preliminary assessment, Nucl. Eng. Des. 340 (2018) 17-30. https://doi.org/10.1016/j.nucengdes.2018.09.023
  14. J. Weisman, B.S. Pei, Prediction of critical heat flux in flow boiling at low qualities, Int. J. Heat Mass Tran. 26 (10) (1983) 1463-1476. https://doi.org/10.1016/S0017-9310(83)80047-7
  15. C.H. Lee, I. Mudawar, A mechanistic critical heat flux model for subcooled flow boiling based on local bulk conditions, Int. J. Multipliase Flow 14 (6) (1988) 711-728. https://doi.org/10.1016/0301-9322(88)90070-5
  16. W.S. Lin, C.H. Lee, B.S. Pei, An improved theoretical critical heat flux model for low quality flow, Nucl. Technol. 88 (1989) 294-306. https://doi.org/10.13182/NT89-A34312
  17. Y. Katto, A prediction model of subcooled water flow boiling CHF for pressure in the range 0.1-20 MPa, Int. J. Heat Mass Tran. 35 (5) (1992) 1115-1123. https://doi.org/10.1016/0017-9310(92)90172-O
  18. D. Bestion, et al., Review of available data for validation of nuresim two-phase CFD software applied to CHF investigations, Sci. Technol. Nucl. Install. (2008), https://doi.org/10.1155/2009/214512.
  19. X. Cheng, U. Muller, Review on Critical Heat Flux in Water Cooled Reactors, FZKA, Karlsruhe, 2003.
  20. B. Niceno, Y. Sato, A. Badillo, et al., Multi-scale modeling and analysis of convective boiling: towards the prediction of CHF in rod bundles, Nucl. Eng. Technol. 42 (6) (2010) 620-635. https://doi.org/10.5516/NET.2010.42.6.620
  21. Wei Liu, Shinian Peng, Guangming Jiang, et al., Development and assessment of a new rod-bundle CHF correlation for China fuel assemblies, Ann. Nucl. Energy 138 (2020) ID107175.
  22. F. de Crecy, About critical heat flux correlations using inlet conditions, in: 31th Meeting of the European Two-phase Flow Group, 1994. Piacenza, Italy.
  23. Y. Haramura, Y. Katto, A new hydrodynamic model of critical heat flux, applicable widely to both pool and forced convention boiling on submerged bodies in saturated liquids, Int. J. Heat Mass Tran. 26 (1983) 389-399. https://doi.org/10.1016/0017-9310(83)90043-1

Cited by

  1. Investigation on Rod Bundle CHF Mechanistic Model for DNB and DO Prediction Under Wide Parameter Range vol.9, 2021, https://doi.org/10.3389/fenrg.2021.620970