Abstract: The sub channel analysis method can consider the mass, energy and momentum conservation process of each channel in the reactor assembly in detail, and has higher computational efficiency than the CFD method. So the sub channel analysis method is the major approach for the thermal hydraulic safety analyze of sodium cooled fast reactor. In order to calculate and analyze the thermal hydraulic characteristics of sodium cooled fast reactor assembly under uniform and inclined power distribution conditions, in this paper, a sub channel code SPLICA for analysis of sodium cooled fast reactor assembly was developed by applying the tow region wire warp mixing model. The mixing effect between channels wound by wire warp warp was considered in detail through the two region model constitutive relationship in SPLICA code. In the two region model, the whole assembly was divided into two regions, including the inner region composed of internal channels and the outer region composed of edge channels and corner channels. In this model, the mixing effect of the wound wire in the inner region and the outer region was considered in different ways. In the inner region, the effect of the wire warp was considered as enhancement of the diffusion between channels. In the outer region, because the wire warp passed through the peripheral gap between the fuel rod and the assembly box in the same direction every time, it brings unidirectional convective mixing in the peripheral gap. And the object oriented C++ language was used to complete the development of the sub channel analysis code SPLICA. In SPLICA code, the energy conservation equation was implicitly solved by SOR iterative method on each axial plane of the assembly, and the momentum conservation equation and mass conservation equation were combined and solved by axial isobaric approximation method. The SPLICA code could accept system pressure, inlet flow or component bundle pressure drop, inlet coolant temperature or coolant specific enthalpy as its boundary conditions. The SPLICA code was validated by the experimental data of FFM 2A (fuel failure mockup) and WARD (Westinghouse advanced reactors division) 61 rod bundle. By comparing with the temperature distribution data of FFM 2A 19 rod bundle experiment and WARD 61 rod bundle experiment, it is verified that this sub channel program developed in this paper has good applicability for the thermal hydraulic analysis of sodium cooled fast reactor components under the conditions of laminar flow, turbulence and laminar turbulence transition region, and the calculation results also have high accuracy under the condition of radial power inclination of fuel assembly. For FFM 2A experiment, compared with the COBRA Ⅳ code, the sub channel analysis code developed in this paper is in better agreement with the experimental values. For the WARD 61 rod bundle inclined power distribution experiment, the temperature distribution calculated by Cheng Todreas mixing model in the laminar turbulent transition zone is in better agreement with the experiment than Zhukov mixing model. This code can provide an effective design and analysis tool for the thermal hydraulic research of pool sodium cooled fast reactor assembly.