Experimental results from the unprotected phase of the Natural Convection (NC) Test performed in the Phenix reactor in the frame of end-of-life (EOL) experiments have been used to qualify the FAST code system for Sodium-cooled Fast Reactor (SFR) transient analysis. An appropriate point kinetics model has been developed with the TRACE code. This has first been used for carrying out a sensitivity study of the core reactivity to the safety coefficients and fuel parameters. Thereby, the main reactivity feedback mechanisms involved in the NC test have been identified, an accurate modeling of the fuel behavior being found to be necessary for the adequate simulation of the core axial expansion. The developed model has been shown to satisfactorily predict the experimental core power and reactivity evolution during the considered transient. Further analysis performed with a 3D neutron-kinetics coupled TRACE/PARCS model has enabled identification of certain current limitations of the FAST code system. Nevertheless, for the very first time, it has been possible to assess the quality of the coupled computational tool on the basis of an actual fast reactor transient. The study thus represents an important step towards the further validation of the FAST code system. (C) 2012 Elsevier Ltd. All rights reserved.