The tremendous advancement in computer technology makes it feasible to perform high-fidelity and multi-physics numerical simulations of reactors, allowing for a more accurate and comprehensive description of the phenomena that occur within nuclear reactors. The application of the high-fidelity coupled neutronic and thermal-hydraulic codes to execute the safety analysis and assessment of the fresh highly enriched uranium core of the IAEA-10 MW MTR research reactor is presented in this research. A three-dimensional coupled neutronic and thermal-hydraulic code is realized using the open-source C++ library OpenFOAM. The coupled code used in this study utilizes an internal coupling approach, where the identical program and grid system are utilized to solve both the neutron physical and thermal-hydraulic fields. The neutron physical field in the reactor is solved by the three-dimensional multi-group time-dependent neutron diffusion equations, and the CFD methodology is applied in solving the thermal-hydraulic governing equations. The global and local steady-state calculation results of MTR are obtained, which are in accordance with the results issued from the IAEA TECDOC and other previous relevant studies. The unprotected loss of flow accident transient condition is considered to conduct the simulation and analysis of the MTR research reactor. The evolutions of temporal and spatial distributions of important parameters of the MTR research reactor are acquired under the ULOFA transient condition, revealing the responses of the reactor core during the transient accident without SCRAM. The results obtained show that the high-fidelity N-TH coupling code can be utilized to analyze the safety of the MTR research reactor.