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Remove basic and boxes examples
This commit is contained in:
Paul Romano
parent
24f7048494
commit
eeb7925376
10 changed files with 0 additions and 408 deletions
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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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import numpy as np
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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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fuel1 = openmc.Material(material_id=1, name='fuel')
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fuel1.set_density('g/cc', 4.5)
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fuel1.add_nuclide('U235', 1.)
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fuel2 = openmc.Material(material_id=2, name='depleted fuel')
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fuel2.set_density('g/cc', 4.5)
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fuel2.add_nuclide('U238', 1.)
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moderator = openmc.Material(material_id=3, 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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# Instantiate a Materials collection and export to XML
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materials_file = openmc.Materials([fuel1, fuel2, moderator])
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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 planar surfaces
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x1 = openmc.XPlane(surface_id=1, x0=-10)
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x2 = openmc.XPlane(surface_id=2, x0=-7)
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x3 = openmc.XPlane(surface_id=3, x0=-4)
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x4 = openmc.XPlane(surface_id=4, x0=4)
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x5 = openmc.XPlane(surface_id=5, x0=7)
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x6 = openmc.XPlane(surface_id=6, x0=10)
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y1 = openmc.YPlane(surface_id=11, y0=-10)
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y2 = openmc.YPlane(surface_id=12, y0=-7)
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y3 = openmc.YPlane(surface_id=13, y0=-4)
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y4 = openmc.YPlane(surface_id=14, y0=4)
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y5 = openmc.YPlane(surface_id=15, y0=7)
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y6 = openmc.YPlane(surface_id=16, y0=10)
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z1 = openmc.ZPlane(surface_id=21, z0=-10)
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z2 = openmc.ZPlane(surface_id=22, z0=-7)
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z3 = openmc.ZPlane(surface_id=23, z0=-4)
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z4 = openmc.ZPlane(surface_id=24, z0=4)
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z5 = openmc.ZPlane(surface_id=25, z0=7)
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z6 = openmc.ZPlane(surface_id=26, z0=10)
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# Set vacuum boundary conditions on outside
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for surface in [x1, x6, y1, y6, z1, z6]:
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surface.boundary_type = 'vacuum'
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# Instantiate Cells
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inner_box = openmc.Cell(cell_id=1, name='inner box')
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middle_box = openmc.Cell(cell_id=2, name='middle box')
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outer_box = openmc.Cell(cell_id=3, name='outer box')
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# Use each set of six planes to create solid cube regions. We can then use these
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# to create cubic shells.
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inner_cube = +x3 & -x4 & +y3 & -y4 & +z3 & -z4
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middle_cube = +x2 & -x5 & +y2 & -y5 & +z2 & -z5
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outer_cube = +x1 & -x6 & +y1 & -y6 & +z1 & -z6
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outside_inner_cube = -x3 | +x4 | -y3 | +y4 | -z3 | +z4
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# Use surface half-spaces to define regions
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inner_box.region = inner_cube
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middle_box.region = middle_cube & outside_inner_cube
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outer_box.region = outer_cube & ~middle_cube
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# Register Materials with Cells
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inner_box.fill = fuel1
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middle_box.fill = fuel2
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outer_box.fill = moderator
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# Instantiate root universe
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root = openmc.Universe(universe_id=0, name='root universe')
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root.add_cells([inner_box, middle_box, outer_box])
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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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uniform_dist = openmc.stats.Box(*outer_cube.bounding_box, 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 plots.xml File
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###############################################################################
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plot = openmc.Plot(plot_id=1)
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plot.origin = [0, 0, 0]
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plot.width = [20, 20]
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plot.pixels = [200, 200]
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plot.color_by = 'cell'
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# Instantiate a Plots collection and export to XML
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plot_file = openmc.Plots([plot])
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plot_file.export_to_xml()
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@ -1,15 +0,0 @@
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<?xml version="1.0"?>
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<geometry>
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<!-- Definition of Cells -->
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<cell id="1" universe="0" fill="37" region="-2" />
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<cell id="100" universe="37" material="40" region="-1" />
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<cell id="101" universe="37" material="41" region="1" />
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<cell id="2" universe="0" material="41" region="2 -3" />
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<!-- Defition of Surfaces -->
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<surface id="1" type="z-cylinder" coeffs="0 0 7" />
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<surface id="2" type="z-cylinder" coeffs="0 0 9" />
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<surface id="3" type="z-cylinder" coeffs="0 0 11" boundary="vacuum" />
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</geometry>
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<?xml version="1.0"?>
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<materials>
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<material id="40">
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<density value="4.5" units="g/cc" />
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<nuclide name="U235" ao="1.0" />
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</material>
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<material id="41">
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<density value="1.0" units="g/cc" />
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<nuclide name="H1" ao="2.0" />
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<nuclide name="O16" ao="1.0" />
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<sab name="c_H_in_H2O"/>
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</material>
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</materials>
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<?xml version="1.0"?>
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<settings>
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<run_mode>eigenvalue</run_mode>
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<batches>15</batches>
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<inactive>5</inactive>
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<particles>10000</particles>
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<!-- Starting source -->
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<source>
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<space type="box">
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<parameters>-4 -4 -4 4 4 4</parameters>
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</space>
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</source>
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</settings>
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<?xml version="1.0"?>
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<tallies>
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<filter id="1" type="cell">
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<bins>100</bins>
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</filter>
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<filter id="2" type="energy">
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<bins>0 20.0e6</bins>
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</filter>
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<filter id="3" type="energyout">
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<bins>0 20.0e6</bins>
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</filter>
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<tally id="1">
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<filters>1</filters>
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<scores>total scatter nu-scatter absorption fission nu-fission</scores>
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</tally>
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<tally id="2">
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<filters>1 2</filters>
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<scores>total scatter nu-scatter absorption fission nu-fission</scores>
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</tally>
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<tally id="3">
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<filters>1 2 3</filters>
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<scores>scatter nu-scatter nu-fission</scores>
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</tally>
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</tallies>
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<?xml version="1.0"?>
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<geometry>
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<!--
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This example consists of three nested boxes, and is meant to show how to
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use Boolean operators to construct complex cell regions.
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-->
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<surface id="1" type="x-plane" coeffs="-10" boundary="vacuum" />
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<surface id="2" type="x-plane" coeffs="-7" />
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<surface id="3" type="x-plane" coeffs="-4" />
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<surface id="4" type="x-plane" coeffs="4" />
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<surface id="5" type="x-plane" coeffs="7" />
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<surface id="6" type="x-plane" coeffs="10" boundary="vacuum" />
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<surface id="11" type="y-plane" coeffs="-10" boundary="vacuum" />
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<surface id="12" type="y-plane" coeffs="-7" />
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<surface id="13" type="y-plane" coeffs="-4" />
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<surface id="14" type="y-plane" coeffs="4" />
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<surface id="15" type="y-plane" coeffs="7" />
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<surface id="16" type="y-plane" coeffs="10" boundary="vacuum" />
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<surface id="21" type="z-plane" coeffs="-10" boundary="vacuum" />
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<surface id="22" type="z-plane" coeffs="-7" />
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<surface id="23" type="z-plane" coeffs="-4" />
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<surface id="24" type="z-plane" coeffs="4" />
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<surface id="25" type="z-plane" coeffs="7" />
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<surface id="26" type="z-plane" coeffs="10" boundary="vacuum" />
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<!-- Innermost cube -->
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<cell id="1" material="1" region="3 -4 13 -14 23 -24" />
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<!-- Middle cubic shell -->
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<cell id="2" material="2" region="2 -5 12 -15 22 -25 (-3 | 4 | -13 | 14 | -23 | 24)" />
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<!-- Outermost cubic shell -->
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<cell id="3" material="3" region="1 -6 11 -16 21 -26 ~(2 -5 12 -15 22 -25)" />
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</geometry>
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@ -1,21 +0,0 @@
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<?xml version="1.0"?>
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<materials>
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<material id="1">
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<density value="4.5" units="g/cc" />
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<nuclide name="U235" ao="1.0" />
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</material>
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<material id="2">
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<density value="4.5" units="g/cc" />
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<nuclide name="U238" ao="1.0" />
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</material>
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<material id="3">
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<density value="1.0" units="g/cc" />
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<nuclide name="O16" ao="1.0" />
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<nuclide name="H1" ao="2.0" />
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<sab name="c_H_in_H2O" />
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</material>
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</materials>
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@ -1,9 +0,0 @@
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<?xml version="1.0"?>
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<plots>
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<plot id="1" type="slice">
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<color_by>cell</color_by>
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<origin>0. 0. 0.</origin>
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<width>20. 20.</width>
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<pixels>200 200</pixels>
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</plot>
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</plots>
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<?xml version="1.0"?>
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<settings>
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<!-- Parameters for k-eigenvalue calculation -->
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<run_mode>eigenvalue</run_mode>
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<batches>15</batches>
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<inactive>5</inactive>
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<particles>10000</particles>
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<!-- Starting source -->
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<source>
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<space type="box" parameters="-10. -10. -10. 10. 10. 10." />
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</source>
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</settings>
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