Assessment of the 3-D Thermal-Hydraulic Nuclear Core Computer Code FLICA-IV on Rod Bundle Experiments

Abstract

The physical validation compared with the hydraulic and two-phase flow experiments of the thermal-hydraulic FLICA-IV nuclear core computer code, in the case of a pressurized water reactor is presented. This three-dimensional two-phase flow code is devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores. The four balance equations used by the code and the closure relationships are first presented. Then, the facilities employed for the code validation are described. They are the ones that use either laser velocimetry techniques in the case of hydraulic validation to measure accurately the flow field around rods or isokinetic sampling to carry out the qualities and the axial mass velocities at the outlet of a rod bundle in the case of two-phase flow validation. Comparisons between experimental and computed values are then presented for the axial flow blockage simulation, inlet assemblies flow mixing, axial flow spacer grid disturbance, and an outlet rod bundle map of qualities and axial mass velocities.