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132 lines
4.5 KiB
Python
132 lines
4.5 KiB
Python
import openmc
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from openmc.source import Source
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from openmc.stats import Box
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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 = 15
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inactive = 5
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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 some Nuclides
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h1 = openmc.Nuclide('H-1')
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o16 = openmc.Nuclide('O-16')
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u235 = openmc.Nuclide('U-235')
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# Instantiate some Materials and register the appropriate Nuclides
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moderator = openmc.Material(material_id=41, name='moderator')
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moderator.set_density('g/cc', 1.0)
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moderator.add_nuclide(h1, 2.)
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moderator.add_nuclide(o16, 1.)
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moderator.add_s_alpha_beta('HH2O', '71t')
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fuel = openmc.Material(material_id=40, 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 all Materials, 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_materials([moderator, 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 ZCylinder surfaces
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surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=7, name='surf 1')
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surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=9, name='surf 2')
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surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=11, name='surf 3')
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surf3.boundary_type = 'vacuum'
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='cell 1')
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cell2 = openmc.Cell(cell_id=100, name='cell 2')
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cell3 = openmc.Cell(cell_id=101, name='cell 3')
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cell4 = openmc.Cell(cell_id=2, name='cell 4')
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# Use surface half-spaces to define regions
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cell1.region = -surf2
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cell2.region = -surf1
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cell3.region = +surf1
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cell4.region = +surf2 & -surf3
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# Register Materials with Cells
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cell2.fill = fuel
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cell3.fill = moderator
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cell4.fill = moderator
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# Instantiate Universes
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universe1 = openmc.Universe(universe_id=37)
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root = openmc.Universe(universe_id=0, name='root universe')
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cell1.fill = universe1
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# Register Cells with Universes
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universe1.add_cells([cell2, cell3])
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root.add_cells([cell1, cell4])
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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.source = Source(space=Box([-4, -4, -4], [4, 4, 4]))
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settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC tallies.xml File
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###############################################################################
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# Instantiate some tally Filters
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cell_filter = openmc.Filter(type='cell', bins=100)
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energy_filter = openmc.Filter(type='energy', bins=[0., 20.])
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energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.])
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# Instantiate the first Tally
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first_tally = openmc.Tally(tally_id=1, name='first tally')
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first_tally.filters = [cell_filter]
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scores = ['total', 'scatter', 'nu-scatter',
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'absorption', 'fission', 'nu-fission']
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first_tally.scores = scores
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# Instantiate the second Tally
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second_tally = openmc.Tally(tally_id=2, name='second tally')
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second_tally.filters = [cell_filter, energy_filter]
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second_tally.scores = scores
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# Instantiate the third Tally
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third_tally = openmc.Tally(tally_id=3, name='third tally')
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third_tally.filters = [cell_filter, energy_filter, energyout_filter]
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third_tally.scores = ['scatter', 'nu-scatter', 'nu-fission']
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# Instantiate a TalliesFile, register all Tallies, and export to XML
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tallies_file = openmc.TalliesFile()
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tallies_file.add_tally(first_tally)
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tallies_file.add_tally(second_tally)
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tallies_file.add_tally(third_tally)
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tallies_file.export_to_xml()
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