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https://github.com/openmc-dev/openmc.git
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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>
91 lines
3 KiB
Python
91 lines
3 KiB
Python
from pathlib import Path
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import pytest
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import numpy as np
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import openmc
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from openmc.utility_funcs import change_directory
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from tests.testing_harness import HashedPyAPITestHarness
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def build_model(shared_secondary):
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openmc.reset_auto_ids()
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# Material
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w = openmc.Material(name='Tungsten')
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w.add_element('W', 1.0)
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w.set_density('g/cm3', 19.25)
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materials = openmc.Materials([w])
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# Geometry surfaces
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x0 = openmc.XPlane(x0=0.0, boundary_type='reflective')
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x1 = openmc.XPlane(x0=160.0, boundary_type='vacuum')
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y0 = openmc.YPlane(y0=0.0, boundary_type='reflective')
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y1 = openmc.YPlane(y0=160.0, boundary_type='reflective')
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z0 = openmc.ZPlane(z0=0.0, boundary_type='reflective')
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z1 = openmc.ZPlane(z0=160.0, boundary_type='reflective')
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region = +x0 & -x1 & +y0 & -y1 & +z0 & -z1
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cell = openmc.Cell(region=region, fill=w)
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root = openmc.Universe(cells=[cell])
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geometry = openmc.Geometry(root)
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# Source: planar on x=0, mono-directional along +x, 14.1 MeV neutrons
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space = openmc.stats.CartesianIndependent(
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openmc.stats.Discrete([0.01], [1.0]),
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openmc.stats.Uniform(0.0, 160.0),
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openmc.stats.Uniform(0.0, 160.0),
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)
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angle = openmc.stats.Monodirectional((1.0, 0.0, 0.0))
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energy = openmc.stats.Discrete([14.1e6], [1.0])
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source = openmc.IndependentSource(space=space, angle=angle, energy=energy)
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settings = openmc.Settings()
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settings.run_mode = 'fixed source'
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settings.batches = 5
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settings.particles = 50
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settings.source = source
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settings.shared_secondary_bank = shared_secondary
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model = openmc.Model(geometry=geometry, materials=materials, settings=settings)
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# Mesh tally: 1 cm voxels, flux only
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mesh = openmc.RegularMesh()
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mesh.dimension = (20, 20, 1)
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mesh.lower_left = (0.0, 0.0, 0.0)
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mesh.upper_right = (160.0, 160.0, 160.0)
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mesh_filter = openmc.MeshFilter(mesh)
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flux_tally = openmc.Tally(name='flux')
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flux_tally.filters = [mesh_filter]
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flux_tally.scores = ['flux']
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tallies = openmc.Tallies([flux_tally])
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model.tallies = tallies
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parent_dir = Path(__file__).parent
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lower_ww_bounds = np.loadtxt(parent_dir / 'ww_n.txt')
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weight_windows = openmc.WeightWindows(mesh,
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lower_ww_bounds,
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upper_bound_ratio=5.0,
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particle_type='neutron')
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model.settings.weight_windows = weight_windows
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model.settings.weight_window_checkpoints = {'surface': True,
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'collision': True}
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model.settings.survival_biasing = True
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return model
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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_weight_windows_with_survival_biasing(shared_secondary, subdir):
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with change_directory(subdir):
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model = build_model(shared_secondary)
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harness = HashedPyAPITestHarness('statepoint.5.h5', model)
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harness.main()
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