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https://github.com/openmc-dev/openmc.git
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126 lines
4.4 KiB
Python
126 lines
4.4 KiB
Python
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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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