The legitimate distribution channel for MCNP6.2 (and later official releases) is the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory. The MCNP project team at Los Alamos provides information about requesting the code and on installation, documentation, and training on the official MCNP website. RSICC supplies executable-only packages (and source-code packages in some cases) for supported platforms, accompanying documentation, verification/validation problems, and ACE-formatted nuclear data libraries.
OpenMC is a Monte Carlo particle transport code developed at MIT. It is open-source (MIT license), free to download, and actively maintained. mcnp62 download free
If your work specifically requires MCNP6.2 (e.g., for regulatory compliance or specific nuclear data libraries), you must go through the official channels. The legitimate distribution channel for MCNP6
For Users in the United States: You must contact the RSICC (Radiation Safety Information Computational Center) at Oak Ridge National Laboratory. For International Users: The process is more rigorous
For International Users: The process is more rigorous due to export controls.