In the course of severe accident in a nuclear power plant, the corium mixture may flow down in the reactor pit and start a thermal attack of the basemat concrete. A simple way to terminate the melt progression is to add water on top of it. Among the various physical processes that may participate in the quenching, water may penetrate into the solidified corium through cracks generated by the thermal stress during the solidification. The present paper aims at providing a clarification of the process though an analysis of the few experimental data available using proto-typical corium, namely the SSWICS 1-7 experiments. At first, a general reminder and analysis of the data is given. Then, the thermal-hydraulic aspects are investigated through the construction of a dedicated heat flux correlation and its use via the experimentally measured post-test permeabilities. The analysis is supported by 1D and 2D evaluations with the multi-phase flow code MC3D-PREMIX, slightly modified for the purpose. It is concluded that the heat flux drastically decreases with the amount of added concrete material. Furthermore, 2D border effects are investigated and their importance highlighted in view of the experimental results analysed. These effects should explain the absence of observation of water ingression heat flux in the cases with large concrete amounts. Following, the paper proposes a model for the created permeability. Due to the complexity of the process and to the large uncertainties of the needed material properties in the considered situation with very high temperatures, a semi-empirical model is derived and reproduces the available data to a good approximation. Lastly, the model is adapted to the situation with internal decay heat, although no open experimental data is available to precisely support any model. A complete modeling is out of the scope of the paper, hence the focus is to provide hints for a first analysis of the impact of residual power on the melt progression and its coolability.