[en] Simulations of new reactor designs, such as generation IV concepts, require three dimensional modeling to ensure a sufficiently realistic description for safety analysis. If precise solutions of local physical phenomena (DNBR, cross flow, form factors,...) are to be found then the use of accurate 3D coupled neutronics/thermal-hydraulics codes becomes essential. Moreover, to describe this coupled field with a high level of accuracy requires successive iterations between neutronics and thermal-hydraulics at equilibrium until convergence (power deposits and temperatures must be finely discretized, ex: pin by pin and axial discretization). In this paper we present the development and simulation results of such coupling capabilities using our code MURE (MCNP Utility for Reactor Evolution), a precision code written in C++ which automates the preparation and computation of successive MCNP calculations either for precision burnup and/or thermal-hydraulics/thermic purposes. For the thermal-hydraulics part, the code COBRA is used. It is a sub-channel code that allows steady-state and transient analysis of reactor cores. The goal is a generic, non system-specific code, for both burn-up calculations and safety analysis at any point in the fuel cycle: the eventual trajectory of an accident scenario will be sensitive to the initial distribution of fissile material and neutron poisons in the reactor (axial and radial heterogeneity). The MURE code is open-source, portable and manages all the neutronics and the thermal-hydraulics/thermic calculations in background: control is provided by the MURE interface or the user can interact directly with the codes if desired. MURE automatically builds input files and other necessary data, launches the codes and manages the communication between them. Consequently accurate 3D simulations of power plants on both global and pin level of detail with thermal feedback can be easily performed (radial and axial meshing grids are managed by MURE). A comparison to an NEA benchmark of a heterogeneous PWR MOX/UO₂ core is presented. Results for hot full-power conditions show an agreement of our simulations with the benchmark (the accuracy of the results are within the errors of the benchmark). The temperature dependent cross sections for MCNP have been provided for each isotope using NJOY99. The convergence of coupled field in heterogeneous configuration is obtained after around five iterations; the Shannon entropy effect which affects neutron source convergence is attenuated using a large number of source particles and inactive cycles. (authors)