Abstract
We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.
Article PDF
Similar content being viewed by others
Avoid common mistakes on your manuscript.
References
Fast Reactor Database 2006 Update, IAEA Technical Document No. 1531 (IAEA, 2006).
A. V. Zhukov, A. P. Sorokin, and Yu. A. Kuzina, “Emergency cooling down of fast-neutron reactors by natural convection (a review),” Therm. Eng. 60, 345–354 (2013). doi 10.1134/S0040601513050108
D. G. Zaryugin, V. M. Poplavskii, V. I. Rachkov, A. P. Sorokin, Yu. E. Shvetsov, S. A. Rogozhkin, and S. F. Shepelev, “Computational and experimental validation of the planned emergency heat-removal system for BN-1200,” At. Energy 116, 271–277 (2014).
A. N. Opanasenko, V. M. Selivanov, and N. N. Shan’gin, “Effect of unsteady natural convection on the structure of a liquid flow in the horizontal mixing chamber,” Inzh.-Fiz. Zh. 39, 603–610 (1980).
J. J. Lorenz and P. A. Howard, “Entrainment by a jet at a density interface in a thermally stratified vessel,” J. Heat Transfer 101, 538–542 (1979).
D. Tenchine, “Some thermal hydraulic challenges in sodium cooled fast reactors,” Nucl. Eng. Des. 240, 1195–1217 (2010).
D. G. Zaryugin, S. G. Kalyakin, S. T. Leskin, A. N. Opanasenko, and A. P. Sorokin, “Numerical and experimental investigations of thermalhydraulic characteristics for fast reactor vessals on integral model SARH in different operation regimes,” Izv. Vyssh. Uchebn. Zaved., Yad. Energ., No. 2, 96–104 (2013).
D. Weinberg, K. Rust, and H. Hoffmann, Overview Report of RAMONA-NEPTUN Program on Passive Decay Heat Removal (Forschungszentrum Karlsruhe, 1996). http://bibliothek.fzk.de/zb/berichte/FZKA5667.pdf.
D. Wilhelm, G. Hansen, and H. Strotmann, “The 1: 10 scale chain of natural convection “KIWA” for studying the LMR passive decay heat removal,” in Proc. Int. Conf. on Design and Safety of Advanced Nuclear Power Plants (ANP’92), Tokyo, Japan, Oct. 25–29, 1992 (At. Energy Soc. of Japan, 1992), Vol. 3, Paper 26.4.
V. V. Pakholkov, A. I. Potseluev, S. A. Rogozhkin, and S. F. Shepelev, “Experimental studies of BN reactor cooling modes on a thermal-hydraulic test facility “TISEY”,” in Proc. 18th Int. Conf. for Young Specialists on Nuclear Power Plants (Gidropress, Podolsk, 2016). http://www.gidropress.podolsk.ru/files/proceedings/kms-2016/documents/kms2016-035.pdf.
D. A. Sergeev, “A measuring system for studying liquid flows by the particle image velocimetry method based on a diode-pumped solid-state laser,” Instrum. Exp. Tech. 52, 438–444 (2009).
D. A. Sergeev, A. A. Kandaurov, Yu. I. Troitskaya, V. V. Pakholkov, S. A. Rogozhkin, and S. F. Shepelev, “A particle-image velocimetry system for measurement of velocity flow fields for investigations of thermohydraulic processes on the large-scale benchmark of a promising fast-neutron reactor,” Instrum. Exp. Tech. 60, 418–427 (2017).
Author information
Authors and Affiliations
Corresponding author
Additional information
Original Russian Text © V.V. Pakholkov, A.A. Kandaurov, A.I. Potseluev, S.A. Rogozhkin, D.A. Sergeev, Yu.I. Troitskaya, S.F. Shepelev, 2017, published in Teploenergetika.
Rights and permissions
About this article
Cite this article
Pakholkov, V.V., Kandaurov, A.A., Potseluev, A.I. et al. Experimental investigation of a new method for advanced fast reactor shutdown cooling. Therm. Eng. 64, 496–503 (2017). https://doi.org/10.1134/S0040601517070059
Received:
Accepted:
Published:
Issue Date:
DOI: https://doi.org/10.1134/S0040601517070059