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141 lines
4.8 KiB
Python
141 lines
4.8 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 = 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.setDensity('g/cc', 1.0)
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moderator.addNuclide(h1, 2.)
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moderator.addNuclide(o16, 1.)
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moderator.addSAlphaBeta('HH2O', '71t')
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fuel = openmc.Material(material_id=40, name='fuel')
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fuel.setDensity('g/cc', 4.5)
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fuel.addNuclide(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.setDefaultXS('71c')
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materials_file.addMaterials([moderator, fuel])
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materials_file.exportToXML()
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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.setBoundaryType('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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# Register Surfaces with Cells
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cell1.addSurface(surface=surf2, halfspace=-1)
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cell2.addSurface(surface=surf1, halfspace=-1)
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cell3.addSurface(surface=surf1, halfspace=+1)
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cell4.addSurface(surface=surf2, halfspace=+1)
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cell4.addSurface(surface=surf3, halfspace=-1)
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# Register Materials with Cells
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cell2.setFill(fuel)
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cell3.setFill(moderator)
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cell4.setFill(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.setFill(universe1)
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# Register Cells with Universes
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universe1.addCells([cell2, cell3])
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root.addCells([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.setRootUniverse(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.setGeometry(geometry)
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geometry_file.exportToXML()
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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.setBatches(batches)
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settings_file.setInactive(inactive)
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settings_file.setParticles(particles)
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settings_file.setSourceSpace('box', [-4, -4, -4, 4, 4, 4])
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settings_file.exportToXML()
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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, label='first tally')
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first_tally.addFilter(cell_filter)
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scores = ['total', 'scatter', 'nu-scatter', \
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'absorption', 'fission', 'nu-fission']
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for score in scores:
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first_tally.addScore(score)
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# Instantiate the second Tally
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second_tally = openmc.Tally(tally_id=2, label='second tally')
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second_tally.addFilter(cell_filter)
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second_tally.addFilter(energy_filter)
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scores = ['total', 'scatter', 'nu-scatter', \
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'absorption', 'fission', 'nu-fission']
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for score in scores:
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second_tally.addScore(score)
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# Instantiate the third Tally
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third_tally = openmc.Tally(tally_id=3, label='third tally')
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third_tally.addFilter(cell_filter)
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third_tally.addFilter(energy_filter)
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third_tally.addFilter(energyout_filter)
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scores = ['scatter', 'nu-scatter', 'nu-fission']
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for score in scores:
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third_tally.addScore(score)
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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.addTally(first_tally)
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tallies_file.addTally(second_tally)
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tallies_file.addTally(third_tally)
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tallies_file.exportToXML()
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