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121 lines
4.2 KiB
Python
121 lines
4.2 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 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_element('H', 2.)
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moderator.add_element('O', 1.)
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moderator.add_s_alpha_beta('c_H_in_H2O')
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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 Materials collection and export to XML
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materials_file = openmc.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, register the root Universe, and export to XML
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geometry = openmc.Geometry(root)
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geometry.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 Settings object, set all runtime parameters, and export to XML
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settings_file = openmc.Settings()
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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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# Create an initial uniform spatial source distribution over fissionable zones
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bounds = [-4., -4., -4., 4., 4., 4.]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
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settings_file.source = openmc.source.Source(space=uniform_dist)
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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.CellFilter(cell2)
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energy_filter = openmc.EnergyFilter([0., 20.e6])
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energyout_filter = openmc.EnergyoutFilter([0., 20.e6])
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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 Tallies collection and export to XML
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tallies_file = openmc.Tallies((first_tally, second_tally, third_tally))
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tallies_file.export_to_xml()
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