Abstract: In order to assess transient responses and parameters of sodium cooled fast reactor (SFR) under all type of accidents whether satisfy the acceptance criteria, and validate the reliability and feasibility of computational fluid dynamics model when applying lower Prandtl fluids, a CFD method substituting for traditional system analysis was adopted to simulate loss of flow without scram experiment for fast flux test facility (FFTF), which is a loop type SFR. The threedimensional model of FFTF was constructed, which includes inlet plenum, annular plenum, invessel storage, peripheral plenum, core basket, hot pool, and core. It should be noted that the fuel assemblies were simplified into porous mediums while the inter wrapper was reserved, and the barriers of pool were simplified into plate without thickness, the standard kε turbulence model was adopted. In the simulation process, two procedures are adopted, i.e., steady debug and transient analysis. In the steady debug, the porous parameters of fuel assemblies were adjusted by code, for the wire region, the empirical relations were adopted and kept uncharged, for inlet region, because the local form resistance is always hard to get, the porous parameters of those regions were adjusted by comparing the calculated mass flow rate with design mass flow rate until the error was within acceptance range. In the transient analysis step, the loss of flow without scram #13 (LOFWOS#13) was selected to simulate, and the simulation time lasts 900 s. The numerical results were compared with experiments and the local phenomena of FFTF during accidents process were analyzed. For the comparison between numerical results and experiments, the numerical results of outlet temperature have a little difference between experiments results for the assembly which installed thermocouple. For the local phenomena, the CFD method could capture well for the complex thermal hydraulic phenomena of hot pool such as thermal stratification, and the temperature changes only 4 K, which imply that the heat capacity of the huge pool has large thermal-inertia which is a positive influence for safety of SFR. And from the temperature distribution of core, the location of hot channel in the core is changed versus time, and the inter wrapper flow is highly dependent on core temperature distribution, and under the lower power and lower mass flow rate condition, the inter wrapper could flat core temperature distribution. Those phenomena could not be tracked by traditional system code, and show that it is necessity to adopt three dimensional CFD method in the safety analysis. However, due to small difference between experiments results and lack of enough validation of CFD methods, there are still to develop accurate model and be validated by much more experiment especially full reactor experiments to satisfy tools requirements in security review process for nuclear regulatory.