Hello all,
I was wondering if I could get a little clarification on the results of a tally that I’m trying to troubleshoot. I’m trying to find a tally that can create an axial representation of the flux in the x and z dimensions, and while I’m getting results that make sense in the behavior, the actual results don’t make any sense. The chunk of code below is what we’re using to define the tallies:
mesh = openmc.RegularMesh()
mesh_height = 100
mesh_width = mesh_height
mesh.dimension = [mesh_width, 1, mesh_height]
mesh.lower_left = [-200, -200, -200]
mesh.upper_right = [200, 200, 200]
mesh_filter = openmc.MeshFilter(mesh)
tally = openmc.Tally()
tally.filters.append(mesh_filter)
tally.scores = [“flux”]
tallies = openmc.Tallies([tally])
tally_written = tallies.export_to_xml()
I’m curious if there’s some kind of automatic normalization of tally results because the resulting mesh plot when run through our Python script gives a maximum flux value of about 0.57. Considering how we are running at a keff of above 1 I think the flux behavior is being modeled properly in the code but I’m not pulling the proper data.
Any help would be appreciated, and let me know if I can clarify further. I want to scale this exact same tally to fission rate and heating so if any additional input on those types of tallies and needed syntax to run them properly would help as well.
Thanks again!