Abstract: The transient analysis of unprotected accident is significant in safety analysis of sodium-cooled fast reactor. Based on MOX-3600 and MET-1000 benchmarks published by OECD/NEA, SARAX code system was applied to do transient calculation for different sodium-cooled fast reactors. Various reactivity feedback effects in the reactor were analyzed, and the changes of fuel temperature and coolant temperature during unprotected loss of flow (ULOF) transient and unprotected transient over power (UTOP) transient were calculated. The results show that SARAX code system can give reasonable results of parameter prediction in transient analysis of fast reactor. ULOF transient is more severe for sodium-cooled fast reactor in that it will cause the boiling of sodium and then cause severe accident.