I have been using OpenMC for a variety of (usually fixed source) neutron transport problems that can often be simplified down to something with rotational symmetry. Recently I’ve encountered shielding calculations where neutron flux can be reduced by 6 orders of magnitudes or more by neutron reflectors instead of absorbers, so simple survival biasing does not produce great performance improvements. As I’m not aware of any previous problems to adapt OpenMC for 2D axisymmetric problems, I’ve been thinking of improving the statistics on my mesh tallies by aggregating all cells with the same radius to the axis-of-symmetry.
I’d first like to make sure I’ve not missed any previous effort of trying to do this in the past, please point out anything related to this if you know of any!
To be clear, I don’t mean moving fully to 2D transport, but implementing a cylindrical mesh and an aggregate score to average across cells in the phi direction in the Python post-processing code. I plan to start implementing something for my specific problem here using rectilinear and regular meshes soon if there is no obvious existing solution.