Toward an Innovative CFD-Based Upscaling Methodology to Elaborate Closures for the FLICA4 Subchannel Code
Résumé
The physical validation of 3D component scale thermalhydraulic nuclear core codes is so far based either on experimental data, or on sensitivity analyses when no reliable experimental data are available. In the present paper, innovative methodologies based on upscaling from CEA TrioCFD fine scale computations are presented. Within the framework of the 17 17 Pressurized Water Reactor (PWR) square rod bundle design, we present here three kinds of methodologies addressing the upscaling topic, all based on numerical simulations at CFD scale. First, a methodology to establish the friction coefficients in the direction parallel to rods for typical subchannel cells is presented. Secondly, an upscaling of the average axial velocities in a N x N rod bundle without any grid and wrapped in a square box is used to investigate the FLICA4 turbulent viscosity model coefficient. Last, for a generic support grid, the global singular pressure drop coefficient based on a CFD computation on the complete, full cross-section of 17 x 17 assembly, is evaluated. These
three kinds of approach show the potential benefit of upscaling fine scale CFD simulations for component thermalhydraulic nuclear core codes in term of experimental cost saving and investigation of closure relationships
often difficult to address by the classical experimental approach.