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162 lines
5.3 KiB
Python
162 lines
5.3 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 = 20
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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 some Materials and register the appropriate Nuclides
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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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moderator = openmc.Material(material_id=2, 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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iron = openmc.Material(material_id=3, name='iron')
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iron.set_density('g/cc', 7.9)
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iron.add_element('Fe', 1.)
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# Instantiate a Materials collection and export to XML
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materials_file = openmc.Materials([moderator, fuel, iron])
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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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left = openmc.XPlane(surface_id=1, x0=-3, name='left')
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right = openmc.XPlane(surface_id=2, x0=3, name='right')
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bottom = openmc.YPlane(surface_id=3, y0=-4, name='bottom')
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top = openmc.YPlane(surface_id=4, y0=4, name='top')
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fuel_surf = openmc.ZCylinder(surface_id=5, x0=0, y0=0, r=0.4)
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left.boundary_type = 'vacuum'
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right.boundary_type = 'vacuum'
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top.boundary_type = 'vacuum'
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bottom.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=101, name='cell 2')
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cell3 = openmc.Cell(cell_id=102, name='cell 3')
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cell4 = openmc.Cell(cell_id=500, name='cell 4')
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cell5 = openmc.Cell(cell_id=600, name='cell 5')
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cell6 = openmc.Cell(cell_id=601, name='cell 6')
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# Use surface half-spaces to define regions
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cell1.region = +left & -right & +bottom & -top
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cell2.region = -fuel_surf
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cell3.region = +fuel_surf
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cell5.region = -fuel_surf
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cell6.region = +fuel_surf
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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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cell5.fill = iron
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cell6.fill = moderator
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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univ2 = openmc.Universe(universe_id=3)
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univ3 = openmc.Universe(universe_id=4)
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cells with Universe
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univ1.add_cells([cell2, cell3])
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univ2.add_cells([cell4])
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univ3.add_cells([cell5, cell6])
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root.add_cell(cell1)
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# Instantiate a Lattice
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lattice = openmc.HexLattice(lattice_id=5)
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lattice.center = [0., 0., 0.]
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lattice.pitch = [1., 2.]
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lattice.universes = \
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[ [ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ],
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[ [univ2] + [univ1]*11, [univ2] + [univ1]*5, [univ1] ],
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[ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ] ]
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lattice.outer = univ2
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# Fill Cell with the Lattice
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cell1.fill = lattice
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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 = [-1, -1, -1, 1, 1, 1]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
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settings_file.source = openmc.IndependentSource(
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space=uniform_dist, constraints={'fissionable': True})
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settings_file.keff_trigger = {'type' : 'std_dev', 'threshold' : 5E-4}
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settings_file.trigger_active = True
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settings_file.trigger_max_batches = 100
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settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC plots.xml file
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###############################################################################
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plot_xy = openmc.SlicePlot(plot_id=1)
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plot_xy.filename = 'plot_xy'
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plot_xy.origin = [0, 0, 0]
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plot_xy.width = [6, 6]
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plot_xy.pixels = [400, 400]
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plot_xy.color_by = 'material'
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plot_yz = openmc.SlicePlot(plot_id=2)
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plot_yz.filename = 'plot_yz'
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plot_yz.basis = 'yz'
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plot_yz.origin = [0, 0, 0]
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plot_yz.width = [8, 8]
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plot_yz.pixels = [400, 400]
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plot_yz.color_by = 'material'
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# Instantiate a Plots collection, add plots, and export to XML
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plot_file = openmc.Plots((plot_xy, plot_yz))
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plot_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 a distribcell Tally
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tally = openmc.Tally(tally_id=1)
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tally.filters = [openmc.DistribcellFilter(cell2)]
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tally.scores = ['total']
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# Instantiate a Tallies collection and export to XML
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tallies_file = openmc.Tallies([tally])
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tallies_file.export_to_xml()
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