Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Research on void drift between rod bundle subchannels

  • Shasha Liu (Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University) ;
  • Zaiyong Ma (Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University) ;
  • Bo Pang (Department of Nuclear Science and Technology, Shenzhen University) ;
  • Rui Zhang (Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University) ;
  • Luteng Zhang (Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University) ;
  • Quanyao Ren (Science and Technology on Reactor System Design Technology Laborator) ;
  • Liangming Pan (Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University)
  • Received : 2023.11.09
  • Accepted : 2024.03.21
  • Published : 2024.08.25
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Abstract

Void drift between subchannels in a rod bundle is a crucial phenomenon affecting the calculation accuracy of thermal-hydraulic parameters in SMRs. It holds significant importance in enhancing the precision of safety analysis for SMRs. Existing research on experiment and model of void drift between rod bundle subchannels is relatively rare, and the accuracy of model calculations requires improvement. In this study, experiments on gas-liquid two-phase non-equilibrium flow were conducted to measure the redistribution of two-phase flow induced by void drift in a 1 × 2 rod bundle. The experiment results indicated that in bubby flow regime with void fraction less than 0.3, the void diffusion coefficient showed little variation with changes in void fraction. However, in slug flow and annular flow regimes with void fraction exceeding 0.3, the void diffusion coefficient significantly increased with an increase in void fraction. Furthermore, a new void drift model was developed and validated based on a subchannel code. The overall predicted uncertainty for the outlet void fraction in the rod bundle benchmark was less than 13%.

Keywords

Acknowledgement

The authors appreciate the National Natural Science Foundation of China (U21B2059,12105273)

References

  1. Guo Chengzhan, Haige Zhao, Progress of operation and application on Shenzhen MNSR, Chinese Journal of Nuclear Science and Engineering 20 (2000).
  2. Huai-kun Jiang, et al., Application of MNSR epithermal neutron activation analysis in determination of geological sample, Atomic Energy Sci. Technol. 49 (8) (2015) 1488.
  3. Li Yiguo, et al., Physics experimental study for reactor of in-hospital neutron irradiator, Atomic Energy Sci. Technol. 43 (2009).
  4. M.D. Carelli, D.T. Ingersoll, Handbook of Small Modular Nuclear Reactors, 2014.
  5. Chenglong Wang, et al., Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor, Int. J. Energy Res. 42 (8) (2018) 2643-2654.
  6. Hang Xia, et al., Development of a thermo-mechanical coupling code based on an annular fuel sub-channel code, Nucl. Eng. Des. 355 (2019) 110284.
  7. Yonghong Tian, et al., Sub-Channel analysis of Pb-Bi-cooled reactor with modified COBRA-EN, in: International Conference on Nuclear Engineering, 45912, American Society of Mechanical Engineers, 2014.
  8. J.T. Rogers, N.E. Todreas, Coolant Interchannel Mixing in Reactor Fuel Rod Bundles, 1969.
  9. R.T. Lahey Jr., F.J. Moody, The Thermal-Hydraulics of a Boiling Water Nuclear Reactor, second ed., 1993.
  10. R.T. Lahey, F.A. Schraub, Mixing, Flow Regimes, and Void Fraction for Two-phase Flow in Rod Bundles, 1969.
  11. R.T. Lahey, et al., Out-of-pile subchannel measurements in a nine-rod bundle for water at 1000 psia, in: Proceedings of the International Symposium on Two-phase Systems 6, 1972, pp. 345-363.
  12. Jose M. Gonzalez-Santalo, Two Phase Flow Mixing in Rod Bundle Subchannels, Diss. Massachusetts Institute of Technology, 1971.
  13. Jr Lahey, T. Richard, Subchannel measurements of the equilibrium quality and mass flux distribution in a rod bundle, in: International Heat Transfer Conference Digital Library, Begel House Inc., 1986.
  14. Y. Sato, M. Sadatomi, H. Tsukashima, Two-phase flow characteristics in interconnected subchannels with different cross-sectional areas, in: Proc. 1987 ASME.JSME Therm. Eng. Joint Cont., Honolulu 5, 1987, pp. 389-395.
  15. A. Tapucu, et al., Experimental investigation of mass exchanges between two laterally interconnected two-phase flows, Nucl. Eng. Des. 105 (3) (1988) 295-312.
  16. Sarman Gencay, Alberto Teyssedou, Peter Tye, Lateral mixing mechanisms in vertical and horizontal interconnected subchannel two-phase flows, Nucl. Technol. 138 (2) (2002) 140-161.
  17. Michio Sadatomi, Akimaro Kawahara, Yoshifusa Sato, Flow redistribution due to void drift in two-phase flow in a multiple channel consisting of two subchannels, Nucl. Eng. Des. 148 (2-3) (1994) 463-474.
  18. Michio Sadatomi, et al., Two-phase void drift phenomena in a 2×3 rod bundle: flow redistribution data and their analysis, Nucl. Technol. 152 (1) (2005) 23-37.
  19. M.P. Sharma, A.K. Nayak, Experimental investigation of void drift in simulated subchannels of a natural-circulation pressure tube-type BWR, Nuclear Science & Engineering the Journal of the American Nuclear Society (2017) 1-12.
  20. A. Kawahara, M. Sadatomi, K. Kano, et al., Void diffusion coefficient in two-phase void drift for several channels of two-and multi-subchannel systems, Multiphas. Sci. Technol. 18 (1) (2006).
  21. Akimaro Kawahara, et al., Void diffusion coefficient in two-phase void drift for several channels of two-and multi-subchannel systems, Multiphas. Sci. Technol. 18 (2006) 1.
  22. Ulrich Bieder, Francois Falk, Gauthier Fauchet, LES analysis of the flow in a simplified PWR assembly with mixing grid, Prog. Nucl. Energy 75 (2014) 15-24.
  23. I.I.I.L.D. Smith, et al., Benchmarking computational fluid dynamics for application to PWR fuel, in: International Conference on Nuclear Engineering, 35979, 2002.
  24. Moyses A. Navarro, AC Santos Andre, Evaluation of a numeric procedure for flow simulation of a 5×5 PWR rod bundle with a mixing vane spacer, Prog. Nucl. Energy 53 (8) (2011) 1190-1196.
  25. Aleksandr Aleksandrovich Armand, The resistance during the movement of a two-phase system in horizontal pipes, Izv. Vses. Teplotekh. Inst. 1 (1946) 16-23.
  26. Salomon Levy, Forced convection subcooled boiling-prediction of vapor volumetric fraction, Int. J. Heat Mass Tran. 10 (7) (1967) 951-965.
  27. K.H. Sun, R.B. Duffey, C.M. Peng, The prediction of two-phase mixture level and hydrodynamically-controlled dryout under low flow conditions, Int. J. Multiphas. Flow 7 (5) (1981) 521-543.
  28. S.G. Beus, Two-phase Turbulent Mixing Model for Flow in Rod Bundles, No. WAPD-T-2438; CONF-711112-6, Bettis Atomic Power Lab., Pittsburgh, Pa, 1972.