Abstract: The nuclear reactor core is a highly heterogeneous system where various physical phenomena are interrelated and coupled. The complex coupled interaction necessitates multi-physics calculations to realize more accurate and realistic simulations in core analysis, especially for the transient conditions of fast reactors. A full-core three-dimensional multi-physics calculation method was researched based on the fast reactor code system MOSASAUR, which is focused on both core steady-state and transient analyses of fast reactor. The deterministic two-step calculation strategy based on the homogenization theory is utilized in MOSASAUR to perform the reactor core neutronics analysis. There are four main functional modules in the previous version: cross-sections generation module, flux spectrum correction module, core simulation module and sensitivity and uncertainty analysis module. In cross-sections generation module, the collision probability method was used to determine the neutron flux of the typical assemblies. Without using the flux spectrum correction module, the neutron flux for the equivalent one-dimensional assembly would be calculated to collapse the cross-sections from 1 968 groups into 33 groups. In the homogenization process, the super homogenization method was optional to be used as the homogenization technique. Based on the 33-group cross-sections, core simulation module simulated core neutron behaviors based on the neutron transport solvers, which was S_N method with triangular grid. The subchannel code COBRA-YT was utilized to carry out the thermal hydraulic calculations. By using COBRA-YT, the distributions of channel coolant temperature and fuel temperature were updated for the thermal feedback calculation of MOSASAUR. In order to extend the capability of COBRA-YT for LFR simulations and analyses, a series of modifications and enhancements have been implemented, including integrating thermophysical properties and empirical correlations related to liquid metals. By the modified COBRA-YT, the temperature distributions of fuel and coolant, pressure drop and coolant flow of LFR will be determined for the core steady-state and transient analyses. The stiffness confinement method (SCM) was employed to solve the time-dependent multi-group neutron transport equation. And the MOSASAUR code was modified to expand the capabilities for neutronic kinetics simulation. For the neutron-thermal hydraulic coupling calculation of fast reactor, the thermal feedback module was cooperated with the neutron transport calculation to update the coupling parameters. In the process of the iteration, the temperature distributions of fuel and coolant will be determined through thermal hydraulic calculation, which requires the core power distribution as the input information. The homogenized cross-sections will be updated by the interpolation method, which takes the temperature as the core-state variable. The homogenized cross-sections would be prepared in advance. Based on the updated cross-sections, the core simulation of LFR will be re-calculated and the obtained power distribution will be transferred to the next thermal hydraulic calculation. The LMW benchmark and the problem based on ORAL 19-rod bundle were employed to verify the accuracy of transient calculation and thermal hydraulic module respectively. Finally, the multi-physics calculation method was utilized to simulate the transient process of the MicroURANUS core. The simulation results demonstrate the capability of the constructed multi-physics calculation framework to accurately simulate the transient behaviors of LMRs. Numerical results show the good accuracy of the newly-developed multi-physics calculation module with MOSASAUR.