Numerical and experimental research on nuclear safety is in the end dedicated to understand, on a plant scale, the fundamental physical phenomena which are associated to specific accident scenarios. Hence, the results derived from single effect experiments or reduced scale analysis have to be extrapolated to plant scale whereas plant scale experiments should be evaluated with respect to their applicability to the physics of the specific scenario. For several years, IRSN and CEA have used Computational Fluid Dynamics (CFD) codes for detailed nuclear safety analyses on plant scale. The paper presents a procedure which has been used to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French Pressurized Water Reactor (PWR) in accidental conditions (inherent dilution problem). A ROCOM experiment as well as an UPTF Tram-C3 experiment has been used for this purpose.