The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO2-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.