Physical Modeling of Hydrodynamic and Heat Transfer Processes in Liquid-Metal Cooled Nuclear Power Facilities

The article presents the results from analyzing the application conditions of the thermophysical processes similarity theory as applied to modeling the hydrodynamics and heat transfer in liquid-metal cooled nuclear power facilities, namely, in channels, in the reactor core rod systems, and in the reactor pressure vessel under different operating conditions. It is shown that direct modeling can be used without limitations only for processes whose determined similarity numbers (criteria) are functions of only the system’s geometrical simplexes and one determining criterion. The availability of two determining criteria in the heat transfer description, e.g., the Reynolds and Prandtl numbers, noticeably complicates the modeling. With three determining criteria, direct modeling is impracticable as a rule. In such cases, systematic multivariate experiments must be set up. The aim of such experiments is to reveal the effects that are allowed by the general mathematical model but which cannot be simulated—either analytically or numerically—using state-of-the-art mathematical technologies. It has been shown from numerical and theoretical investigations and from generalization of experiments, including data on temperature distribution patterns in a liquid-metal flow, that the thermal resistance at the coolant–heat-transfer surface interface boundary is essentially zero if the concentration of impurities in the coolant does not exceed their solubility at the circulating metal temperature. In the case of using liquid metals and alloys (Pb, Pb–Bi, Hg, Na, Na–K, Li, etc.), the heat transfer is described under such conditions by a unified dimensionless dependence on the Peclet number. The transfer of heat in fuel assemblies takes place mainly by convective heat transfer, and the temperature field is governed by the increase in liquid-metal temperature. The temperature distribution depends on the classical similarity criteria, including the Reynolds, Peclet (Prandtl), and Grashof criteria, and on the design and thermophysical characteristics of fuel elements and fuel assemblies (the latter serve as the fuel element approximate similarity criterion). Forced circulation in the reactor vessel is simulated in small-scale water models using the Froude and Peclet numbers, and natural circulation is simulated using the Euler number. The similarity of currents in stably stratified coolant zones is determined by the Froude and Peclet numbers and by the local-gradient Richardson number.