Hello everyone,
I would like to start by saying that I’m a new user of OpenMC.
I wanted to try generating a tally for the fission source (nu-fission) and, at the same time, distinguish in which cell the neutrons were generated in a reactor with a hexagonal lattice. Since, as far as I understand, there is currently no way to define a hexagonal mesh in OpenMC, I tried to define a cell for each assembly of the reactor. This way, I could subsequently define the “CellFilter” and “CellBornFilter” filters and then use them in a tally, using “nu-fission” as the score.
However, I noticed that, when using this combination of filters, the bin values are all equal to 0.0. I tried using the CellFilter individually, and it seems to generate the tally correctly, but when I combine it with CellBornFilter, it doesn’t work. Additionally, I also tried combining CellFilter with CellFromFilter, and it works properly.
I might be missing something, but I’m wondering if there’s a potential issue with the CellBornFilter, or if I’m not using it correctly.
Here attached some lines code where I define the Tally.
assembly_cell_filters = openmc.CellFilter(bins=range(5000,5050))
assembly_cell_born_filters = openmc.CellBornFilter(bins=range(5000,5050))
tally_fission_matrix_hexagonal = openmc.Tally(name ="CellFilter and CellBornFilter hexagonal tally")
tally_fission_matrix_hexagonal.filters = [assembly_cell_filters,assembly_cell_born_filters]
tally_fission_matrix_hexagonal.scores = ['nu-fission']
tallies = openmc.Tallies([tally_fission_matrix_hexagonal])
tallies.export_to_xml()
Thank you in advance for your support!