I am looking to perform what is essentially a neutron radiation shielding study, and I am wanting to get some advice as to which simulation program is the best to use. In this study, I am looking to model how DT generated neutrons are attenuated by different shielding components. I have significant experience with other radiation transport simulators (Geant4), but have not used MCNP or OpenMC before. My pros and cons are essentially that while MCNP is the industry standard, it is export-controlled and, assuming I can get access (I believe I can), I would need to teach myself. OpenMC on the other hand is open source with a large community, meaning I believe it would be easier to pick up, but idk how it compares in terms of accuracy. Also, would one look significantly better on a resume? I am assuming MCNP would, but that's just a guess. Are there any other pros/cons that I am overlooking here? Thanks! submitted by /u/El_Grande_Papi [link] [comments]