Hi!
I am trying to tally neutron flux over my geometry with a point source in the origin using a cylindrical mesh, but this generates negative flux values. Has anyone else experienced this? Running the exact same code but with RegularMesh yields believable results.
Running on: OpenMC version 0.15.0
Some snippets from my code:
mesh_detector = omc.CylindricalMesh.from_domain(domain=my_geometry, dimension=[10,10,10])
mesh_filter_detector = omc.MeshFilter(mesh_detector)
cell_filter_det = omc.CellFilter(detector_cell)
particle_filter = omc.ParticleFilter(‘neutron’)
flux_tally = omc.Tally(name=“flux tally”)
flux_tally.filters = [mesh_filter_detector, particle_filter]
flux_tally.scores = [“flux”]
tallies = omc.Tallies([flux_tally])
tallies.export_to_xml()