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92 lines
3 KiB
Python
92 lines
3 KiB
Python
import openmc
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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# OpenMC simulation parameters
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batches = 500
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inactive = 10
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particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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###############################################################################
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# Instantiate a Nuclides
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u235 = openmc.Nuclide('U-235')
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# Instantiate a Material and register the Nuclide
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fuel = openmc.Material(material_id=1, name='fuel')
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fuel.set_density('g/cc', 4.5)
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fuel.add_nuclide(u235, 1.)
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# Instantiate a MaterialsFile, register Material, and export to XML
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materials_file = openmc.MaterialsFile()
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materials_file.default_xs = '71c'
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materials_file.add_material(fuel)
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materials_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC geometry.xml File
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###############################################################################
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# Instantiate Surfaces
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surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
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surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
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surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
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surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
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surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
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surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
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surf1.boundary_type = 'vacuum'
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surf2.boundary_type = 'vacuum'
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surf3.boundary_type = 'reflective'
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surf4.boundary_type = 'reflective'
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surf5.boundary_type = 'reflective'
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surf6.boundary_type = 'reflective'
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# Instantiate Cell
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cell = openmc.Cell(cell_id=1, name='cell 1')
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# Register Surfaces with Cell
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cell.add_surface(surface=surf1, halfspace=+1)
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cell.add_surface(surface=surf2, halfspace=-1)
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cell.add_surface(surface=surf3, halfspace=+1)
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cell.add_surface(surface=surf4, halfspace=-1)
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cell.add_surface(surface=surf5, halfspace=+1)
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cell.add_surface(surface=surf6, halfspace=-1)
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# Register Material with Cell
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cell.fill = fuel
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# Instantiate Universes
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cell with Universe
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root.add_cell(cell)
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.root_universe = root
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# Instantiate a GeometryFile, register Geometry, and export to XML
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geometry_file = openmc.GeometryFile()
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geometry_file.geometry = geometry
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geometry_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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###############################################################################
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# Instantiate a SettingsFile, set all runtime parameters, and export to XML
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settings_file = openmc.SettingsFile()
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
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settings_file.export_to_xml()
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