Use of the CFD-Code CONV-3D in Reactor Applications

Abstract

The results of testing the CONV-3D code and methods of direct numerical modeling, using supercomputers, of turbulent flows in the elements of nuclear power plants are presented for problems as close as possible to real processes: natural and forced convection of sodium in the upper plenums of the Monju nuclear power plant in Japan and mixing of the two flows at different temperatures in the secondary loop of the Phénix reactor in France. The particularities of the method for calculating the thermo- and hydrodynamic characteristics of the equipment in nuclear power plants are listed.