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Experimental validation of simulating natural circulation of liquid metal using water

  • Lee, Min Ho (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Jerng, Dong Wook (School of Energy Systems Engineering, Chung Ang Univ.) ;
  • Bang, In Cheol (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST))
  • Received : 2019.12.03
  • Accepted : 2020.03.07
  • Published : 2020.09.25

Abstract

Liquid metal-cooled reactors use various passive safety systems driven by natural circulation. Investigating these safety systems experimentally is more advantageous by using a simulant. Although numerous experimental approaches have been applied to natural circulation-driven passive safety systems using simulants, there has been no clear validation of the similarity law. To validate the similarity law experimentally, SINCRO-V experiment was conducted using Wood's metal and water for simulant of the Wood's metal. A pair of SINCRO-V facilities with length-scale ratio of 14.1:1 for identical Bo' was investigated, which was the main similarity parameter in temperature field simulation. In the experimental range of 0.2-1.0% of decay heat, the temperature distribution characteristics of the small water facility were very similar to that of the large Wood's metal facility. The temperature of the Wood's metal predicted by the water experiment showed good agreement with the actual Wood's metal temperature. Despite some error factors like discordance of Gr' and property change along the temperature, the water experiment predicted the Wood's metal temperature with an error of 27%. The validity of the similarity law was confirmed by the SINCRO-V experiments.

References

  1. E.E. Feldman, D. Mohr, L.K. Chang, H.P. Planchon, E.M. Dean, P.R. Betten, EBR-II unprotected loss-of-heat-sink predictions and preliminary test results, Nucl. Eng. Des. 101 (1987) 57-66.
  2. H.P. Planchon, J.I. Sackett, G.H. Golden, R.H. Sevy, Implications of EBR-II inherent safety demonstration test, Nucl. Eng. Des. 101 (1987) 75-90.
  3. W.K. Lehto, R.M. Fryer, E.M. Dean, J.F. Koenig, L.K. Chang, D. Mohr, E.E. Feldman, Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II, Nucl. Eng. Des. 101 (1987) 35-44.
  4. F. Yamada, Y. Fukano, H. Nishi,M. Konomura, Development of natural circulation analyticalmodel in super-COPD code and evaluation of core cooling capability in monju during a station blackout, Nucl. Technol. 188 (2014) 292-321.
  5. T. Ishizu, H. Endo, Y. Shindo, K. Haga, An evaluation of passive safety features of the Japanese prototype LMFBR monju, in: Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11), Avignon, France, 2005.
  6. M. Sadawa, H. Arikawa, N. Mizoo, Experiment and analysis on natural convection characteristics in the experimental fast reactor joyo, Nucl. Eng. Des. 120 (1990) 341-347.
  7. J. Yoo, J. Chang, J.Y. Lim, J.S. Cheon, T.H. Lee, S.K. Kim, K.L. Lee, H.K. Joo, Overall system description and safety characteristics of prototype gen IV sodium cooled fast reactor in Korea, Nucl. Eng. Technol. 48 (2016), 10595-11070.
  8. K.L. Lee, K.S. Ha, J.H. Jeong, C.W. Choi, T.K. Jeong, S.J. Ahn, S.W. Lee, W.P. Chang, S.H. Kang, J.Y. Yoo, A preliminary safety analysis for the prototype gen-IV sodium-cooled fast reactor, Nucl. Eng. Technol. 48 (2016) 1071-1082.
  9. E. Hourcade, F. Curnier, T. Mihara, B. Farges, J.F. Dirat, A. Ide, ASTRID Nuclear Island Design: advances in French-Japanese joint team development of decay heat removal systems, in: Proceedings of the 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016), San Francisco, CA, 2016.
  10. C.F. Smith, W.G. Halsey, N.W. Brown, J.J. Sienicki, A. Moisseytsev, D.C. Wade, SSTAR: the US lead-cooled fast reactor (LFR), J. Nucl. Mater. 376 (2008) 255-259.
  11. J. Wallenius, E. Suvdantsetseg, A. Fokau, ELECTRA: European lead-cooled training reactor, Nucl. Technol. 117 (2012) 303-313.
  12. H.A. Abderrahim, P. D'hondt, MYRRHA: a European experimental ads for R&D Applications status at mid-2005 and prospective towards implementation, J. Nucl. Sci. Technol. 44 (2007) 491-498.
  13. N. Tanaka, S. Moriya, S. Ushijima, T. Koga, Y. Eguchi, Prediction method for thermal stratification in a reactor vessel, Nucl. Eng. Des. 120 (1990) 395-402.
  14. Y. Ieda, H. Kamide, H. Ohshima, S. Sugawara, H. Ninokata, Strategy of experimental studies in PNC on natural convection decay heat removal, in: Proceedings of IAEA-IWGFR Specialists' Meeting on "Evaluation of Decay Heat Removal by Natural Convection", O-arai, Japan, 1993.
  15. A. Ono, H. Kamide, J. Kobayashi, N. Doda, Osamu Watanabe, An experimental study on natural circulation decay heat removal system for a loop type fast reactor, J. Nucl. Sci. Technol. 53 (2016) 1385-1396.
  16. G. Coccoluto, P. Gaggini, V. Labanti, M. Tarantino, W. Ambrosini, N. Forgione, A. Napoli, F. Oriolo, Heavy liquid metal natural circulation in a onedimensional loop, Nucl. Eng. Des. 241 (2011) 1301-1309.
  17. C.C. Yue, L.L. Chen, K.F. Lyu, Y. Li, S. Gao, Y.J. Liu, Q.Y. Huang, Flow characteristics of natural circulation in a leadebismuth eutectic loop, Nucl. Sci. Technol. 28 (2017) 28-39.
  18. K.H. Ryu, B.M. Ban, T.H. Lee, J.H. Lee, S.H. Lee, J.H. Cho, S.H. Ko, J.H. Kim, Natural circulation characteristics under various conditions on heavy liquid metal test loop, Int. J. Therm. Sci. 132 (2018) 316-321.
  19. S. Grewal, E. Gluekler, Water simulation of sodium reactors, Chem. Eng. Commun. 17 (1982), 3443-4360.
  20. Y. Eguchi, H. Takeda, T. Koga, N. Tanaka, K. Yamamoto, Quantitative prediction of natural circulation in an LMFR with a similarity law and a water test, Nucl. Eng. Des. 178 (1997) 295-307.
  21. H. Hoffman, D. Weinberg, Y. Ieda, K. Marten, H. Tschoke, H.H. Frey, Kurt Dres, Thermohydraulic investigations of decay heat removal systems by natural convection for liquid-metal fast breeder reactors, Nucl. Technol. 88 (1989) 75-88.
  22. H. Hoffman, D. Wienberg, R. Webstar, Investigation on natural convection decay heat removal for the EFR - status of the Program, in: Proceedings of IAEA-IWGFR Specialists' Meeting on "Passive and Active Safety Features of LMFRs", Oarai, Japan, 1991.
  23. D. Wienberg, H. Hoffman, H. Ohira, G. Schnetgoke, The status study using RAMONA and NEPTUN models on decay heat removal by natural convection for the European fast reactor, in: Proceedings of IAEA-IWGFR Specialists' Meeting on "Evaluation of Decay Heat Removal by Natural Convection in Fast Reactor", Mito, Japan, 1993.
  24. D. Wienberg, K. Rust, H. Hoffmann, Overview Report of RAMONA-NEPTUN Program on Passive Decay Heat Removal, Report FZKA 5667, Forschungszentrum Karlsruhe, 1996.
  25. M. Akutsu, Y. Okabe, K. Satoh, H. Kamide, K. Hayashi, N. Naohara, K. Iwashige, Y. Shibata, Study of thermal-hydraulic characteristics during decay heat removal in a pool-type fast breeder reactor, Nucl. Technol. 98 (1992) 14-26.
  26. H. Takeda, T. Koga, Study on similarity rule for natural circulation water test of LMFBR, in: Proceedings of IAEA-IWGFR Specialists' Meeting on "Evaluation of Decay Heat Removal by Natural Convection in Fast Reactor", Mito, Japan, 1993.
  27. H. Takeda, T. Koga, O. Watanabe, Experimental and computational simulation for natural circulation in an LMFBR, Nucl. Eng. Des. 140 (1993) 331-340.
  28. K. Rust, H. Tschoke, D. Weinberg, Influence of the position and number of decay heat exchangers on the thermal hydraulics of a slab test facility: a comparison of analytical and experimental data, Exp. Therm. Fluid Sci. 9 (1994) 413-425.
  29. V.M. Mente, G.K. Pandey, I. Banerjee, S. Ajesh Kumar, G. Padmakumar, K.K. Rajan, Experimental studies in water for safety grade decay heat removal of prototype fast breeder reactor, Ann. Nucl. Energy 65 (2014) 114-121.
  30. T. Murakami, Y. Eguchi, K. Oyama, O. Watanabe, Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor, Nucl. Eng. Des. 288 (2015) 220-231.
  31. A. Ono, A. Kurihara, M. Tanaka, H. Oshima, H. Kamide, Study on Reactor Vessel Coolability of Sodium-Cooled Fast Reactor under Severe Accident Condition - Water Experiments Using a Scale Model -, ICAPP 2017, Fukui and Kyoto, Japan, 2017.
  32. P. Planquarta, K. van Tichelen, Experimental investigation of accidental scenarios using a scale water model of a HLM reactor, Nucl. Eng. Des. 346 (2019) 10-16.