Thermal hydraulics computational codes used for the design studies and safety assessment of sodium fast reactors need to be able to model a large scope of scales and physical phenomena. IAEA’s Coordinated Research Project on the Fast Flux Test Facility (FFTF) sets a landmark for the validation of such codes by providing precious experimental data on a reactor scale. The transient under consideration is the Loss Of Flow WithOut Scram (LOFWOS) in which the primary circuit undergoes a forced to natural convection transition after the primary pumps trip. In order to predict this transient correctly, thermos-hydraulic codes need to take into account various time and dimension scales as well as interaction with other physics (such as neutronics and thermo-mechanics). Neutronic feedbacks, power distribution, Pellet Cladding Mechanical Interaction (PCMI) and fuel thermal behavior have been calculated based on FFTF’s fuel composition using CEA in-house codes, namely PARIS/ERANOS for neutronics and GERMINAL V2 for thermo-mechanics. Results from these codes are given to CATHARE3 thermal-hydraulics system code. This paper presents the impact of neutronics and themo-mechanics inputs on a system scale calculation.