Thermal Engineering

, Volume 64, Issue 5, pp 336–344 | Cite as

Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

  • A. N. Opanasenko
  • A. P. Sorokin
  • D. G. Zaryugin
  • A. A. Trufanov
Nuclear Power Plants


Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.


fast reactor emergency heat removal system autonomous heat exchanger core upper chamber forced circulation transient regimes emergency cooldown temperature field stratification temperature fluctuations 


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Copyright information

© Pleiades Publishing, Inc. 2017

Authors and Affiliations

  • A. N. Opanasenko
    • 1
  • A. P. Sorokin
    • 1
  • D. G. Zaryugin
    • 1
  • A. A. Trufanov
    • 1
  1. 1.Leypunsky Institute of Physics and Power EngineeringKaluga oblast, ObninskRussia

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