Modeling of natural circulation for the inherent safety analysis of sodium cooled fast reactors
The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor.
The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed.
It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat removal and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor.
The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system’s intermediate circuit and by the heat transfer through the reactor vessel.
The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety.
The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.
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