Hi there!
I would like to ask if OpenMC is capable of simulating shielding problems or it is made for reactor studies only?. E.g: I would like to simulate an Am-Be neutron source and then surround it with layers of different materials and then to tally flux attenuation after each layer to check if these layers shield from neutrons or not. Thanks in advance!
Hi @nuclearsneke. Yes, OpenMC is capable of handling such a problem. You will have to define the neutron spectrum yourself, but otherwise you can run a fixed source problem and study the attenuation in each layer. The most recent release of OpenMC includes the capability to use weight windows, which may be useful for such a study.
I would like to ask whether there can be no fissile material in the fixed source problem?
Fissile material is allowed in fixed source problems. Just be aware that if you add so much that keff > 1, you’ll end up with a runaway chain reaction and OpenMC will crash.
Thanks a lot！I have another question, I want to run an eigenvalue calculation first, then obtain particle information(‘r’, ‘u’, ‘E’ …) on a boundary surface, and then I want to use it as a fixed source to simulate a fixed source problem. Does this approach work? Or is there a way to do it?