Abstract: After the emergency shutdown of sodium cooled fast reactor, the natural circulation of the primary loop through the core is the only way to cool the core fuel. Therefore, it is significant to accurately predict the flow and temperature distribution of sodium flow in the core under the natural circulation conditions to design the passive decay heat removal system of the sodium cooled fast reactor. We used subchannel analysis method and different formula for calculating friction resistance of wire-wrapped assembly to simulate the flow and temperature distribution in a 37 rods fuel assembly. Then we compared the simulation results with experimental data and found that the subchannel code has good accuracy with UCTD formula. As the sodium flow velocity increases, the mix in sodium flow cannot be ignored.