Boiling experiments for the validation of dryout models used in reactor safety

The coolability of fragmented corium is a major issue in reactor safety. Since the long-term coolability of such particle beds is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the pressure field inside the debris has a strong effect on the cooling potential in multi-dimensional cases as expected in severe accidents in light water reactors (LWR). Therefore, the determination of the pressure field for two-phase flows in porous media is one central point of interest.In this context simulation models and in particular dryout models were developed for reactor safety analyses which have to be validated by reliable experimental data. Therefore, basic experimental investigations have been carried out with inductively heated steel balls of 6 or 3 mm diameter to provide a database for the validation and modification of the friction laws included in these dryout models.The performed boiling and dryout experiments show clearly that models without the explicit consideration of the interfacial drag cannot predict the pressure distribution inside a boiling particle bed, not even qualitatively. Against it, models with an explicit consideration of the interfacial drag can describe the distribution of pressure inside a boiling particle bed.