Hi - new to OpenMC and looking to compare point-kernel programs (MicroShield and MegaShield) to a probabilistic approach (similar to MCNP with out the baggage) - 3D plots of dose rates coming from a bucket of nasty isotopes placed in a shielded container for transport. Can OpenMC simulate that kind of environment?
Yes. Its geometry and material definitions are quite similar to those available in MCNP.
Just to confirm. I can model the material in the bucket (actually a fabricated steel bin containing high activity debris/soil) inside a shielded container (steel, concrete, tungsten) and create dose tallies for storage and transport. Can the waste material then be decay modeled to account for storage?
Yes, have a look at our user’s guide section on depletion and transmutation, which would enable you to model the decay of isotopes within the waste container.