So I’m running a simple script with a spherical geometry around 5-6 layers I’m facing a issue where the outer most layers won’t run it either gives 0 for flux or a error in the script and I’m not sure how to resolve this. I’m able to get a flux output for inner layers. I also changed the particles and batches but it doesn’t work.
I have attached my script below
uranium_fuel = openmc.Material(name='Uranium-235')
uranium_fuel.add_element('U', 100, percent_type='ao')
uranium_fuel.set_density('g/cm3', 19.1)
polyethylene = openmc.Material(name='Polyethylene')
polyethylene.add_element('H', 0.143, percent_type='ao')
polyethylene.add_element('C', 0.857, percent_type='ao')
polyethylene.set_density('g/cm3', 0.94)
water = openmc.Material()
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.set_density('g/cm3', 1.0)
lead = openmc.Material(name='Lead')
lead.add_element('Pb', 100, percent_type='wo')
lead.set_density('g/cm3', 11.34)
steel = openmc.Material(name='Steel')
steel.set_density('g/cm3', 7.75)
steel.add_element('Fe', 0.95, percent_type='wo')
steel.add_element('C', 0.05, percent_type='wo')
my_materials = openmc.Materials([uranium_fuel, polyethylene, lead, steel, water])
my_materials.export_to_xml()
vessel_inner = openmc.Sphere(r=290)
fuel_outer_surface = openmc.Sphere(r=340)
first_wall_outer_surface = openmc.Sphere(r=400)
polyethylene_outer_surface = openmc.Sphere(r=450)
lead_outer_surface = openmc.Sphere(r=500)
inner_vessel_region = -vessel_inner
inner_vessel_cell = openmc.Cell(region=inner_vessel_region)
fuel_region = +vessel_inner & -fuel_outer_surface
fuel_cell = openmc.Cell(region=fuel_region)
fuel_cell.fill = uranium_fuel
first_wall_region = +fuel_outer_surface & -first_wall_outer_surface
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = steel
polyethylene_region = +first_wall_outer_surface & -polyethylene_outer_surface
polyethylene_cell = openmc.Cell(region=polyethylene_region)
polyethylene_cell.fill = polyethylene
lead_region = +polyethylene_outer_surface & -lead_outer_surface
lead_cell = openmc.Cell(region=lead_region)
lead_cell.fill = water
my_geometry = openmc.Geometry([inner_vessel_cell, fuel_cell, first_wall_cell, polyethylene_cell, lead_cell])
my_geometry.export_to_xml()
my_settings = openmc.Settings()
my_settings.batches = 2
my_settings.inactive = 0
my_settings.particles = 20000
my_settings.run_mode = 'fixed source'
my_settings.photon_transport = True
my_source = openmc.IndependentSource()
my_source.space = openmc.stats.Point((0, 0, 0))
my_source.angle = openmc.stats.Isotropic()
my_source.energy = openmc.stats.Discrete([2e6], [1])
my_settings.source = my_source
my_settings.export_to_xml()
neutron_particle_filter = openmc.ParticleFilter(['neutron'])
energy_filter = openmc.EnergyFilter(np.linspace(0, 2e6, 719))
shield_cell_filter = openmc.CellFilter(lead_cell)
gamma_flux_shield_tally = openmc.Tally(name='Photon Flux in Shield')
gamma_flux_shield_tally.filters = [shield_cell_filter, neutron_particle_filter, energy_filter]
gamma_flux_shield_tally.scores = ['flux']
shield_heating = openmc.CellFilter(fuel_cell)
gamma_heating = openmc.Tally(name='Heating')
gamma_heating.filters = [shield_heating]
gamma_heating.scores = ['heating']
my_tallies = openmc.Tallies([gamma_flux_shield_tally, gamma_heating])
my_tallies.export_to_xml()