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36 lines
1,017 B
Python
36 lines
1,017 B
Python
import openmc
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# Create a single material
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iron = openmc.Material()
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iron.set_density('g/cm3', 5.0)
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iron.add_element('Fe', 1.0)
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mats = openmc.Materials([iron])
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mats.export_to_xml()
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# Create a 5 cm x 5 cm box filled with iron
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box = openmc.model.RectangularPrism(10.0, 10.0, boundary_type='vacuum')
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cell = openmc.Cell(fill=iron, region=-box)
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geometry = openmc.Geometry([cell])
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geometry.export_to_xml()
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# Tell OpenMC we're going to use our custom source
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settings = openmc.Settings()
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settings.run_mode = 'fixed source'
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settings.batches = 10
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settings.particles = 1000
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source = openmc.CompiledSource()
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source.library = 'build/libsource.so'
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settings.source = source
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settings.export_to_xml()
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# Finally, define a mesh tally so that we can see the resulting flux
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mesh = openmc.RegularMesh()
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mesh.lower_left = (-5.0, -5.0)
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mesh.upper_right = (5.0, 5.0)
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mesh.dimension = (50, 50)
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tally = openmc.Tally()
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tally.filters = [openmc.MeshFilter(mesh)]
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tally.scores = ['flux']
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tallies = openmc.Tallies([tally])
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tallies.export_to_xml()
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