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0169fd9226
Co-authored-by: John Tramm <jtramm@gmail.com> Co-authored-by: Claude Opus 4.6 <noreply@anthropic.com> Co-authored-by: Paul Romano <paul.k.romano@gmail.com> Co-authored-by: Copilot <copilot@github.com>
49 lines
1.7 KiB
Python
49 lines
1.7 KiB
Python
import openmc
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import pytest
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from openmc.utility_funcs import change_directory
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from tests.testing_harness import PyAPITestHarness
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@pytest.mark.parametrize("shared_secondary,subdir", [
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(False, "local"),
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(True, "shared"),
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])
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def test_particle_production_fission(shared_secondary, subdir):
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"""Fixed-source model with fissionable material to test that
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ParticleProductionFilter correctly counts fission-born neutrons,
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with both local and shared secondary bank modes."""
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with change_directory(subdir):
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openmc.reset_auto_ids()
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model = openmc.Model()
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mat = openmc.Material()
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mat.set_density('g/cm3', 18.0)
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mat.add_nuclide('U235', 1.0)
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model.materials.append(mat)
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sph = openmc.Sphere(r=5.0, boundary_type='vacuum')
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cell = openmc.Cell(fill=mat, region=-sph)
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model.geometry = openmc.Geometry([cell])
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source = openmc.IndependentSource()
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source.energy = openmc.stats.delta_function(1.0e6)
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model.settings.particles = 100
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model.settings.run_mode = 'fixed source'
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model.settings.batches = 2
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model.settings.source = source
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model.settings.create_fission_neutrons = True
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model.settings.shared_secondary_bank = shared_secondary
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# ParticleProductionFilter tracking fission neutron production
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ppf = openmc.ParticleProductionFilter(['neutron'])
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neutron_filter = openmc.ParticleFilter(['neutron'])
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tally = openmc.Tally()
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tally.filters = [neutron_filter, ppf]
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tally.scores = ['events']
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tally.estimator = 'analog'
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model.tallies = [tally]
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harness = PyAPITestHarness('statepoint.2.h5', model)
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harness.main()
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