bug report: plot mesh with EnergyFunctionFilter shows zeros everywhere #106

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1. shimwell commented on Jan 12, 2023

   

   shimwell

   on Jan 12, 2023

   MemberAuthor

   More actions

   
2. shimwell commented on Jan 12, 2023

   

   shimwell

   on Jan 12, 2023

   MemberAuthor

   More actions

   minimal code to reproduce

   import openmc
   
   mat_1 = openmc.Material()
   mat_1.add_nuclide('Li6', 1)
   mat_1.set_density('g/cm3', 1)
   my_materials = openmc.Materials([mat_1])
   
   surface_1 = openmc.Sphere(r=100, boundary_type='vacuum')
   region_1 = -surface_1
   cell_1 = openmc.Cell(region=region_1)
   cell_1.fill = mat_1
   
   my_geometry = openmc.Geometry([cell_1])
   
   reg_mesh = openmc.RegularMesh().from_domain(
       my_geometry,
       dimension=[5, 5, 5]
   )
   reg_mesh_filter = openmc.MeshFilter(reg_mesh)
   
   energy_bins_n, dose_coeffs_n = openmc.data.dose_coefficients(
       particle="neutron", geometry="AP"
   )
   energy_function_filter_n = openmc.EnergyFunctionFilter(energy_bins_n, dose_coeffs_n)
   energy_function_filter_n.interpolation == "cubic"
   
   neutron_particle_filter = openmc.ParticleFilter("neutron")
   
   n_reg_mesh_tally = openmc.Tally(name="neutron_dose_on_cell")
   # note that the EnergyFunctionFilter is included as a filter
   n_reg_mesh_tally.filters = [
       neutron_particle_filter,
       energy_function_filter_n,
       reg_mesh_filter
   ]
   n_reg_mesh_tally.id = 1
   n_reg_mesh_tally.scores = ["flux"]
   my_tallies = openmc.Tallies([n_reg_mesh_tally])
   
   my_settings = openmc.Settings()
   
   my_settings.batches = 10
   my_settings.particles = 100000
   my_settings.run_mode = "fixed source"
   
   source = openmc.Source()
   source.angle = openmc.stats.Isotropic()
   source.energy = openmc.stats.Discrete([14e6], [1])
   source.space = openmc.stats.Point((0.0, 0.0, 0.0))
   
   my_settings.source = source
   
   my_model = openmc.Model(settings=my_settings, geometry=my_geometry, tallies=my_tallies, materials=my_materials)
   
   sp_filename = my_model.run()
   
   sp = openmc.StatePoint(sp_filename)
   
   # shows that the tally is not zero
   print(sp.tallies[1].mean)
3. shimwell commented on Jan 12, 2023

   

   shimwell

   on Jan 12, 2023

   Last edited by shimwell

   MemberAuthor

   More actions

   changing this line

   plotter/openmc_plotter/plotmodel.py

   Line 695 in 4f353cc

   data = data[np.array(selected_bins)].sum(axis=0)

   to

    data = data[np.array(selected_bins)].sum(axis=0) [0]

   and this line

   plotter/openmc_plotter/plotmodel.py

   Line 689 in 4f353cc

   if type(tally_filter) == openmc.MeshFilter:

   to

   if type(tally_filter) in [openmc.MeshFilter, openmc.EnergyFunctionFilter]:

   is a temporary fix
4. letibdesneiges commented on Apr 6, 2023

   

   letibdesneiges

   on Apr 6, 2023

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   Hi,

   I have encountered the same problem when trying to plot doserate data from a flux mesh tally with an EnergyFunctionFilter converting flux to doserate.
   All values are displayed as zero. The above solution did not help unfortunately.

   Attached the statePoint file.
   statepoint.200.h5.zip

   openmc-plotter v. 0.3.1 installed via conda on Debian 11.
5. letibdesneiges commented on Apr 7, 2023

   

   letibdesneiges

   on Apr 7, 2023

   More actions

   Implementing @shimwell changes to plotmodel.py and changing:

   plotter/openmc_plotter/plotmodel.py

   Lines 726 to 727 in eeec020

   	# set image data, reverse y-axis
   	image_data = data[::-1, ...]

   to

   # set image data, reverse y-axis
   data=data.squeeze()
   image_data = data[::-1, ...]
   

   solved it for my specific case.
6. kkiesling

   closed this as completedin #114on Apr 10, 2023