The development of Sodium-cooled Fast Reactors (SFRs) is promoted by the Generation IV International Forum (GIF) since its design and operation is competitive against other technologies and projects. In France, considerable experience has been acquired with three experimental facilities: Rapsodie, Phénix and Superphénix. The Superphénix was a large-size SFR that remains as a unique source of data for these cores. Particularly, during the start-up a set of tests were performed to check the safety criteria of the core in which the feedback coefficients of the core were assessed. These feedback coefficients are the K (1°C variation at the inlet of the core temperature), G (1°C variation in the sodium temperature elevation through the core) and H (1% variation of the nominal power). In this paper, the evaluation of these coefficients is performed using two neutronic platforms: ERANOS and the new APOLLO3. By using these two codes the elementary feedback coefficients are calculated and allow evaluating the global feedback coefficients with a simplified model. The results show the same trend than measurements for the K and H coefficients while for the G coefficient a significant discrepancy is observed at 80% nominal power. The use of a thermal-hydraulic system code such as CATHARE-3 (eventual-ly complemented by CFD calculations) is envisaged to better understand the sources of this bias, which might be due to the simplistic thermal hydraulic models unable to tackle some physical effects of the core operation.