Hi All, I’m quite new to OpenMC and I’m attempting to replicate an experiment I have carried out in OpenMC. The experiment is an AmBe neutron source in middle-bottom of a large graphite stack. import openmc # create graphite - 100% carbon, density 2.26g/cm^3 graphite = openmc.Material() graphite.add_element('C', 1.0) graphite.set_density('g/cm3', 2.26) materials = openmc.Materials([graphite]) # cuboid - simple graphite cube for now graphite_surface = openmc.model.RectangularParallelepiped(-70.0, 70.0, -70.0, 70.0, 0, 210.0, boundary_type='vacuum') graphite_cell = openmc.Cell(fill=graphite, region=-graphite_surface) # universe with graphite cell - i think this is right root_universe = openmc.Universe(cells=[graphite_cell]) # geometry - i think this is right geometry = openmc.Geometry(root_universe) #source (AmBe) - confident on this part source = openmc.IndependentSource() source.space = openmc.stats.Point((0,0,10)) source.angle = openmc.stats.Isotropic() source.energy = openmc.stats.Discrete([4.2e6], [1]) #average energy of neutrons from AmBe source is 4.2MeV, want to have a spectrum eventually # settings - this works but not sure if its ideal settings = openmc.Settings() settings.source = source settings.particles = 2500 settings.batches = 20 settings.inactive = 5 settings.output= {'tallies': True} # figure out how to tally neutrons? # export to xml geometry.export_to_xml() materials.export_to_xml() settings.export_to_xml() # run - is this what i want or do I want something else if i'm just trying to get neutron flux/counts openmc.run() Above is what I’ve made so far but not sure how to progress to what I want to get which is ideally a flux measurement throughout the graphite cube Thanks, Nooradin

I have written a small package for plotting regular mesh tallies that might be of interest [image] GitHub - fusion-energy/openmc_regular_mesh_plotter: A Python package for... A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations. - fusion-energy/openmc_regular_mesh_plotter The plotter supports plotting mesh tallies [image] GitHub - openmc-dev/plotter: Native plotting GUI for model design and... Native plotting GUI for model design and verification - openmc-dev/plotter

I’ve got the basics of the code down, I’m still struggling to understand why my tallies are completely empty. I’ve added filters now and some other things to clean it up a bit import openmc import numpy as np # create graphite - 100% carbon, density 2.26g/cm^3 graphite = openmc.Material() graphite.add_element('C', 1.0) graphite.set_density('g/cm3', 2.26) materials = openmc.Materials([graphite]) # cuboid - simple graphite cube for now graphite_surface = openmc.model.RectangularParallelepiped(-70.0, 70.0, -70.0, 70.0, 0, 210.0, boundary_type='vacuum') graphite_cell = openmc.Cell(fill=graphite, region=-graphite_surface) # geometry geometry = openmc.Geometry([graphite_cell]) #source (AmBe) - confident on this part source = openmc.IndependentSource() source.space = openmc.stats.Point((0,0,10)) source.angle = openmc.stats.Isotropic() source.energy = openmc.stats.Discrete([4.2e6], [1]) #average energy of neutrons from AmBe source is 4.2MeV, want to have a spectrum eventually # settings - this works but not sure if its ideal settings = openmc.Settings() settings.source = source settings.run_mode = 'fixed source' settings.particles = 10000 settings.batches = 10 settings.inactive = 5 settings.output = {'tallies': True} # filters neutron_filter = openmc.ParticleFilter(['neutron']) #selects neutrons energy_filter = openmc.EnergyFilter(np.linspace(0, 0.025, 200)) #select thermal neutrons cell_filter = openmc.CellFilter([graphite_cell]) #select cell # tally tally = openmc.Tally(name='FluxTally') tally.scores =['flux'] tally.filters = [cell_filter, neutron_filter, energy_filter] tallies = openmc.Tallies([tally]) # export to xml #geometry.export_to_xml() #materials.export_to_xml() #settings.export_to_xml() #tallies.export_to_xml() #plots.export_to_xml() # run - is this what i want or do I want something else if i'm just trying to get neutron flux/counts model = openmc.model.Model(geometry, materials, settings, tallies) results_filename = model.run(tracks=True) #post process results = openmc.StatePoint(results_filename) cell_tally = results.get_tally(name='FluxTally') Any help on why my tallies are completely empty would be great, thanks!

Are you accessing the cell_tally.mean attribute

I have a tally with results now, however I want to be able to plot a flux “heat map” of a x-z or y-z slice with my data, do you know how I could go about doing that? iimport openmc import numpy as np import matplotlib.pyplot as plt import os # create graphite - 100% carbon, density 2.26g/cm^3 graphite = openmc.Material() graphite.add_element('C', 1.0) graphite.set_density('g/cm3', 2.26) materials = openmc.Materials([graphite]) # cuboid - simple graphite cube for now graphite_surface = openmc.model.RectangularParallelepiped(-70.0, 70.0, -70.0, 70.0, 0, 210.0, boundary_type='vacuum') graphite_cell = openmc.Cell(fill=graphite, region=-graphite_surface) # geometry geometry = openmc.Geometry([graphite_cell]) # source (AmBe) source = openmc.IndependentSource() source.space = openmc.stats.Point((0, 0, 10)) source.angle = openmc.stats.Isotropic() source.energy = openmc.stats.Discrete([4.2e6], [1]) # average energy of neutrons from AmBe source is 4.2MeV, want to have a spectrum eventually # settings settings = openmc.Settings() settings.source = source settings.run_mode = 'fixed source' settings.particles = 1000 settings.batches = 10 settings.inactive = 0 settings.output = {'tallies': True} # filters neutron_filter = openmc.ParticleFilter(['neutron']) # selects neutrons energy_filter = openmc.EnergyFilter(np.linspace(0, 0.025, 201)) # select thermal neutrons energy_filter2 = openmc.EnergyFilter(np.linspace(0, 4.2e6, 201)) # selects all cell_filter = openmc.CellFilter([graphite_cell]) # select cell # mesh filter mesh = openmc.RegularMesh() mesh.dimension = [100, 100] mesh.lower_left = [-70.0, 0] mesh.upper_right = [70.0, 210.0] mesh_filter = openmc.MeshFilter(mesh) mesh_tally = openmc.Tally(name='tallies_on_mesh') mesh_tally.filters = [mesh_filter, energy_filter2, neutron_filter, cell_filter] mesh_tally.scores = ['flux'] tallies = openmc.Tallies([mesh_tally]) if os.path.exists('tallies.h5'): os.remove('tallies.h5') if os.path.exists('statepoint.10.h5'): os.remove('statepoint.10.h5') if os.path.exists('summary.h5'): os.remove('summary.h5') if os.path.exists('tallies.out'): os.remove('tallies.out') if os.path.exists('tracks.h5'): os.remove('tracks.h5') # run model = openmc.model.Model(geometry, materials, settings, tallies) results_filename = model.run() # post process - not sure results = openmc.StatePoint(results_filename) # Get the tally results my_tally = results.get_tally(scores=['flux']) my_slice = my_tally.get_slice(scores=['flux']) print(my_tally) # Get the mean of the tally results mean = my_slice.mean.shape = (mesh.dimension[0], mesh.dimension[1]) # Plot the reshaped tally results plt.imshow(mean, extent=(mesh.lower_left[0], mesh.upper_right[0], mesh.lower_left[1], mesh.upper_right[1])) plt.show() This is what i’ve tried so far but the output is far from ideal. I think the main issue is that my output is one dimensional and I’m not sure what is going wrong with my mesh to make that happen.

Hi, thank you for the packages. I used your first one and I have what seems to be a good slice of data. [graphite_flux] [graphite_flux_log] Is there anyway I could extract the diffusion length from the data produced now. I’ve looked up some sort but it was using multigroup diffusion coefficient but I’m not sure that would be appropiate. Once again really appreciate the help

L^2 should equal 1/6-th of the mean distance between birth and absorption of a thermal neutron. See https://www.nuclear-power.com/nuclear-power/reactor-physics/neutron-diffusion-theory/diffusion-length/ since L=53.5 cm for graphite, i dont think your geometry is big enough to sample the mean ‘life’ distance

I am happy to vary the dimension of the graphite, how do I go about sampling the lifetime of a thermal neutron?

Hi again, Just another question regarding mesh tallies, right now my mesh filter looks like this # mesh filter mesh = openmc.RegularMesh() mesh.dimension = [140, 1, 210] mesh.lower_left = [-70.0, -70.0, 0] mesh.upper_right = [70.0, 70.0, 210.0] mesh_filter = openmc.MeshFilter(mesh) mesh_tally = openmc.Tally(name='mesh_tally') mesh_tally.filters = [mesh_filter, neutron_filter, cell_filter] mesh_tally.scores = ['flux'] tallies = openmc.Tallies([mesh_tally]) Is there a way for the mesh tally to only tally thermal neutrons, I’ve tried adding an energy filter but I get ValueErrors when trying to plot. Thanks

I would have gobe with an energy filter with a couple of low energy values like you mentioned. Not sure why this would cause an error.

Oh a value error with my plotter, oh right yes i can believe that. Can you poster the value error and i shall fix it

ValueError: cannot select an axis to squeeze out which has size not equal to one I’ve tried plotting through another means but it looks quite noisey. [graphite_flux_log_thermal] Had to do some weird resizing with my arrays to get it to work though my_slice = my_tally.get_slice(scores=['flux']) print(my_slice.mean) print(my_slice.mean.shape) new_dim1 = 140 new_dim2 = int(my_slice.mean.shape[0] / new_dim1) reshaped = my_slice.mean.reshape(new_dim1, new_dim2) plt.imshow(reshaped, extent=mesh.bounding_box.extent['xz'], norm=LogNorm()) plt.savefig(f'{material_choice}_flux_log_thermal.png') plt.show() I think this is probably due to do with my energy filters number of bins, not sure what kind of number I should use. energy_filter_thermal = openmc.EnergyFilter(np.linspace(0, 0.625, 200)) # select thermal neutrons energy_filter_all = openmc.EnergyFilter(np.linspace(0, 4.2e6, 200)) # selects all

OpenMC

Neutron flux in a graphite stack

User Support

Shimwell February 29, 2024, 11:06pm 20

Not sure but I wonder if you need a 2nd value in that EnergyFilter, as it is only one value long at the moment. How about putting a 2 at the end

openmc.EnergyFilter(np.linspace(0, 0.030, 2)

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