MCNP is actually the only code in use at ITER for neutron transport modeling. The OpenMC software package is a potential alternative to it. At least, the OpenMC code can be a complement for preliminary evaluations, lost particle elimination, weight calculations for variance reduction, and similar tasks. OpenMC presents powerful and convenient functionality for these tasks. The essential advantage of OpenMC is that it is available as an open-source code. This allows one to adapt the code for one’s specific tasks. The project is supported by more than 100 developers from all around the world. The pace of the code development is excellent because of the experienced team, sound engineering process, and applied modern technologies. This paper discusses the use of the OpenMC code for ITER neutronics modeling, translation and synchronization of the models between MCNP and OpenMC, and some verification results.