CEA has developed Generation IV fast neutron reactors with gas and sodium coolants in the two last tenth of years. Namely these reactor projects were the GFR2400 (gas-cooled fast reactor of 2400 thermal megawatts) and ASTRID (sodium-cooled fast reactor demonstrator of 1500 thermal megawatts). The objective of this paper is to provide a cross-comparison of the severe accident prevention and mitigation capability of these reactor concepts based on the work done during a significant time period and taking benefit of the distance since these studies. This comparison can highlight both generic trends resulting from a common study approach used for both concepts and from very detailed results obtained during their conceptual design studies. Despite their power difference and their fuel element design specificities, the main study results and conclusions are quite generical of the two concepts as explained in the paper. Thus they could be extrapolated to other reactor design providing main reactor features are kept (core materials, coolant, neutron spectrum, etc.). The assessment of core melting prevention relies on the study of the natural behavior of the GFR2400 and of ASTRID when facing the various accident sequence families. Then, the efficiency and the features of the systems to be foreseen for core degradation prevention are presented. As far as mitigation is concerned, all the consequences of core melting are investigated (i.e. the induced loadings in terms of nature and of range) by considering various core degraded states. Based on the magnitude of these loadings, the needs of mitigation means are assessed for each concept. Among other trends, the presented work shows the very good prevention capability of the SFR concept but the necessity to mitigate the fast vaporization and expansion of degraded core materials. Conversely, the limited coolant capability of the GFR concept and its low thermal inertia require a pressurized gas circulation into the core, limiting its prevention capability whereas its core melting should not induce substantial mechanical loadings of the reactor structures. However, for this last concept, thermochemical interactions between the core materials are an issue deeply investigated in order to understand and simulate core degradation.