The thermal-hydraulics challenges of a nuclear reactor are numerous and mastering these is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics (CFD) codes or System Thermal-Hydraulics (STH) codes can address a lot of the different questions, nevertheless the use of water modeling for the study of the thermal-hydraulic behavior of a new primary system and the validation of codes remains an extremely valuable tool. A water model of the pool-type PbBi-cooled MYRRHA reactor has been developed at the von Karman Institute in collaboration with SCK•CEN. It is a full Plexiglas model at a geometrical scale 1/5 of MYRRHA. This transparent water model allows the application of optical measurement techniques, like Particle Image velocimetry (PIV) for the flow characterization. Local results of PIV measurements performed in the lower plenum at the entrance of the core are presented and compared with CFD results for a nominal operating condition and a natural convection case simulating the decay heat removal. A very good agreement has been found in the velocity field. The results also show the importance of the radial flow entering the core of the water model in natural convection.