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adding back files to be reviewed
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Paul Romano
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121
examples/python/basic/build-xml.py
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121
examples/python/basic/build-xml.py
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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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125
examples/python/boxes/build-xml.py
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125
examples/python/boxes/build-xml.py
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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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161
examples/python/lattice/hexagonal/build-xml.py
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161
examples/python/lattice/hexagonal/build-xml.py
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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 = 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:], only_fissionable=True)
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settings_file.source = openmc.source.Source(space=uniform_dist)
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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.Plot(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.Plot(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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169
examples/python/lattice/nested/build-xml.py
Normal file
169
examples/python/lattice/nested/build-xml.py
Normal file
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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 = 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')
|
||||
moderator.set_density('g/cc', 1.0)
|
||||
moderator.add_element('H', 2.)
|
||||
moderator.add_element('O', 1.)
|
||||
moderator.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials((moderator, fuel))
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate Surfaces
|
||||
left = openmc.XPlane(surface_id=1, x0=-2, name='left')
|
||||
right = openmc.XPlane(surface_id=2, x0=2, name='right')
|
||||
bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
|
||||
top = openmc.YPlane(surface_id=4, y0=2, name='top')
|
||||
fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, r=0.4)
|
||||
fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, r=0.3)
|
||||
fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, r=0.2)
|
||||
|
||||
left.boundary_type = 'vacuum'
|
||||
right.boundary_type = 'vacuum'
|
||||
top.boundary_type = 'vacuum'
|
||||
bottom.boundary_type = 'vacuum'
|
||||
|
||||
# Instantiate Cells
|
||||
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
|
||||
cell2 = openmc.Cell(cell_id=2, name='Cell 2')
|
||||
cell3 = openmc.Cell(cell_id=101, name='cell 3')
|
||||
cell4 = openmc.Cell(cell_id=102, name='cell 4')
|
||||
cell5 = openmc.Cell(cell_id=201, name='cell 5')
|
||||
cell6 = openmc.Cell(cell_id=202, name='cell 6')
|
||||
cell7 = openmc.Cell(cell_id=301, name='cell 7')
|
||||
cell8 = openmc.Cell(cell_id=302, name='cell 8')
|
||||
|
||||
# Use surface half-space to define regions
|
||||
cell1.region = +left & -right & +bottom & -top
|
||||
cell2.region = +left & -right & +bottom & -top
|
||||
cell3.region = -fuel1
|
||||
cell4.region = +fuel1
|
||||
cell5.region = -fuel2
|
||||
cell6.region = +fuel2
|
||||
cell7.region = -fuel3
|
||||
cell8.region = +fuel3
|
||||
|
||||
# Register Materials with Cells
|
||||
cell3.fill = fuel
|
||||
cell4.fill = moderator
|
||||
cell5.fill = fuel
|
||||
cell6.fill = moderator
|
||||
cell7.fill = fuel
|
||||
cell8.fill = moderator
|
||||
|
||||
# Instantiate Universe
|
||||
univ1 = openmc.Universe(universe_id=1)
|
||||
univ2 = openmc.Universe(universe_id=2)
|
||||
univ3 = openmc.Universe(universe_id=3)
|
||||
univ4 = openmc.Universe(universe_id=5)
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
univ1.add_cells([cell3, cell4])
|
||||
univ2.add_cells([cell5, cell6])
|
||||
univ3.add_cells([cell7, cell8])
|
||||
root.add_cell(cell1)
|
||||
univ4.add_cell(cell2)
|
||||
|
||||
# Instantiate nested Lattices
|
||||
lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
|
||||
lattice1.lower_left = [-1., -1.]
|
||||
lattice1.pitch = [1., 1.]
|
||||
lattice1.universes = [[univ1, univ2],
|
||||
[univ2, univ3]]
|
||||
|
||||
lattice2 = openmc.RectLattice(lattice_id=6, name='4x4 core')
|
||||
lattice2.lower_left = [-2., -2.]
|
||||
lattice2.pitch = [2., 2.]
|
||||
lattice2.universes = [[univ4, univ4],
|
||||
[univ4, univ4]]
|
||||
|
||||
# Fill Cell with the Lattice
|
||||
cell1.fill = lattice2
|
||||
cell2.fill = lattice1
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-1, -1, -1, 1, 1, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC plots.xml file
|
||||
###############################################################################
|
||||
|
||||
plot = openmc.Plot(plot_id=1)
|
||||
plot.origin = [0, 0, 0]
|
||||
plot.width = [4, 4]
|
||||
plot.pixels = [400, 400]
|
||||
plot.color_by = 'material'
|
||||
|
||||
# Instantiate a Plots object and export to XML
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh(mesh_id=1)
|
||||
mesh.dimension = [4, 4]
|
||||
mesh.lower_left = [-2, -2]
|
||||
mesh.width = [1, 1]
|
||||
|
||||
# Instantiate tally Filter
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1)
|
||||
tally.filters = [mesh_filter]
|
||||
tally.scores = ['total']
|
||||
|
||||
# Instantiate a Tallies collection, register Tally/RegularMesh, and export to
|
||||
# XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
166
examples/python/lattice/simple/build-xml.py
Normal file
166
examples/python/lattice/simple/build-xml.py
Normal file
|
|
@ -0,0 +1,166 @@
|
|||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 20
|
||||
inactive = 10
|
||||
particles = 10000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
fuel = openmc.Material(material_id=1, name='fuel')
|
||||
fuel.set_density('g/cc', 4.5)
|
||||
fuel.add_nuclide('U235', 1.)
|
||||
|
||||
moderator = openmc.Material(material_id=2, name='moderator')
|
||||
moderator.set_density('g/cc', 1.0)
|
||||
moderator.add_element('H', 2.)
|
||||
moderator.add_element('O', 1.)
|
||||
moderator.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([moderator, fuel])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate Surfaces
|
||||
left = openmc.XPlane(surface_id=1, x0=-2, name='left')
|
||||
right = openmc.XPlane(surface_id=2, x0=2, name='right')
|
||||
bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
|
||||
top = openmc.YPlane(surface_id=4, y0=2, name='top')
|
||||
fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, r=0.4)
|
||||
fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, r=0.3)
|
||||
fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, r=0.2)
|
||||
|
||||
left.boundary_type = 'vacuum'
|
||||
right.boundary_type = 'vacuum'
|
||||
top.boundary_type = 'vacuum'
|
||||
bottom.boundary_type = 'vacuum'
|
||||
|
||||
# Instantiate Cells
|
||||
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
|
||||
cell2 = openmc.Cell(cell_id=101, name='cell 2')
|
||||
cell3 = openmc.Cell(cell_id=102, name='cell 3')
|
||||
cell4 = openmc.Cell(cell_id=201, name='cell 4')
|
||||
cell5 = openmc.Cell(cell_id=202, name='cell 5')
|
||||
cell6 = openmc.Cell(cell_id=301, name='cell 6')
|
||||
cell7 = openmc.Cell(cell_id=302, name='cell 7')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
cell1.region = +left & -right & +bottom & -top
|
||||
cell2.region = -fuel1
|
||||
cell3.region = +fuel1
|
||||
cell4.region = -fuel2
|
||||
cell5.region = +fuel2
|
||||
cell6.region = -fuel3
|
||||
cell7.region = +fuel3
|
||||
|
||||
# Register Materials with Cells
|
||||
cell2.fill = fuel
|
||||
cell3.fill = moderator
|
||||
cell4.fill = fuel
|
||||
cell5.fill = moderator
|
||||
cell6.fill = fuel
|
||||
cell7.fill = moderator
|
||||
|
||||
# Instantiate Universe
|
||||
univ1 = openmc.Universe(universe_id=1)
|
||||
univ2 = openmc.Universe(universe_id=2)
|
||||
univ3 = openmc.Universe(universe_id=3)
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
univ1.add_cells([cell2, cell3])
|
||||
univ2.add_cells([cell4, cell5])
|
||||
univ3.add_cells([cell6, cell7])
|
||||
root.add_cell(cell1)
|
||||
|
||||
# Instantiate a Lattice
|
||||
lattice = openmc.RectLattice(lattice_id=5)
|
||||
lattice.lower_left = [-2., -2.]
|
||||
lattice.pitch = [1., 1.]
|
||||
lattice.universes = [[univ1, univ2, univ1, univ2],
|
||||
[univ2, univ3, univ2, univ3],
|
||||
[univ1, univ2, univ1, univ2],
|
||||
[univ2, univ3, univ2, univ3]]
|
||||
|
||||
# Fill Cell with the Lattice
|
||||
cell1.fill = lattice
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-1, -1, -1, 1, 1, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.trigger_active = True
|
||||
settings_file.trigger_max_batches = 100
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC plots.xml file
|
||||
###############################################################################
|
||||
|
||||
plot = openmc.Plot(plot_id=1)
|
||||
plot.origin = [0, 0, 0]
|
||||
plot.width = [4, 4]
|
||||
plot.pixels = [400, 400]
|
||||
plot.color_by = 'material'
|
||||
|
||||
# Instantiate a Plots collection and export to XML
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh(mesh_id=1)
|
||||
mesh.dimension = [4, 4]
|
||||
mesh.lower_left = [-2, -2]
|
||||
mesh.width = [1, 1]
|
||||
|
||||
# Instantiate tally Filter
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate tally Trigger
|
||||
trigger = openmc.Trigger(trigger_type='rel_err', threshold=1E-2)
|
||||
trigger.scores = ['all']
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1)
|
||||
tally.filters = [mesh_filter]
|
||||
tally.scores = ['total']
|
||||
tally.triggers = [trigger]
|
||||
|
||||
# Instantiate a Tallies collection and export to XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
138
examples/python/pincell/build-xml.py
Normal file
138
examples/python/pincell/build-xml.py
Normal file
|
|
@ -0,0 +1,138 @@
|
|||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 100
|
||||
inactive = 10
|
||||
particles = 1000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
|
||||
uo2.set_density('g/cm3', 10.29769)
|
||||
uo2.add_element('U', 1., enrichment=2.4)
|
||||
uo2.add_element('O', 2.)
|
||||
|
||||
helium = openmc.Material(material_id=2, name='Helium for gap')
|
||||
helium.set_density('g/cm3', 0.001598)
|
||||
helium.add_element('He', 2.4044e-4)
|
||||
|
||||
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
|
||||
zircaloy.set_density('g/cm3', 6.55)
|
||||
zircaloy.add_element('Sn', 0.014 , 'wo')
|
||||
zircaloy.add_element('Fe', 0.00165, 'wo')
|
||||
zircaloy.add_element('Cr', 0.001 , 'wo')
|
||||
zircaloy.add_element('Zr', 0.98335, 'wo')
|
||||
|
||||
borated_water = openmc.Material(material_id=4, name='Borated water')
|
||||
borated_water.set_density('g/cm3', 0.740582)
|
||||
borated_water.add_element('B', 4.0e-5)
|
||||
borated_water.add_element('H', 5.0e-2)
|
||||
borated_water.add_element('O', 2.4e-2)
|
||||
borated_water.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='Fuel OR')
|
||||
clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.40005, name='Clad IR')
|
||||
clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=0.45720, name='Clad OR')
|
||||
left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
|
||||
right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
|
||||
bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
|
||||
top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
|
||||
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
gap = openmc.Cell(cell_id=2, name='cell 2')
|
||||
clad = openmc.Cell(cell_id=3, name='cell 3')
|
||||
water = openmc.Cell(cell_id=4, name='cell 4')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
fuel.region = -fuel_or
|
||||
gap.region = +fuel_or & -clad_ir
|
||||
clad.region = +clad_ir & -clad_or
|
||||
water.region = +clad_or & +left & -right & +bottom & -top
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.fill = uo2
|
||||
gap.fill = helium
|
||||
clad.fill = zircaloy
|
||||
water.fill = borated_water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
root.add_cells([fuel, gap, clad, water])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
entropy_mesh = openmc.RegularMesh()
|
||||
entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
|
||||
entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
|
||||
entropy_mesh.dimension = [10, 10, 1]
|
||||
settings_file.entropy_mesh = entropy_mesh
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh()
|
||||
mesh.dimension = [100, 100, 1]
|
||||
mesh.lower_left = [-0.62992, -0.62992, -1.e50]
|
||||
mesh.upper_right = [0.62992, 0.62992, 1.e50]
|
||||
|
||||
# Instantiate some tally Filters
|
||||
energy_filter = openmc.EnergyFilter([0., 4., 20.e6])
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1, name='tally 1')
|
||||
tally.filters = [energy_filter, mesh_filter]
|
||||
tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
|
||||
# Instantiate a Tallies collection and export to XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
49
examples/python/pincell_depletion/chain_simple.xml
Normal file
49
examples/python/pincell_depletion/chain_simple.xml
Normal file
|
|
@ -0,0 +1,49 @@
|
|||
<?xml version="1.0"?>
|
||||
<depletion_chain>
|
||||
<nuclide name="I135" decay_modes="1" reactions="1" half_life="2.36520E+04">
|
||||
<decay type="beta" target="Xe135" branching_ratio="1.0" />
|
||||
<reaction type="(n,gamma)" Q="0.0" target="Xe136" /> <!-- Not precisely true, but whatever -->
|
||||
</nuclide>
|
||||
<nuclide name="Xe135" decay_modes="1" reactions="1" half_life="3.29040E+04">
|
||||
<decay type=" beta" target="Cs135" branching_ratio="1.0" />
|
||||
<reaction type="(n,gamma)" Q="0.0" target="Xe136" />
|
||||
</nuclide>
|
||||
<nuclide name="Xe136" decay_modes="0" reactions="0" />
|
||||
<nuclide name="Cs135" decay_modes="0" reactions="0" />
|
||||
<nuclide name="Gd157" decay_modes="0" reactions="1" >
|
||||
<reaction type="(n,gamma)" Q="0.0" target="Nothing" />
|
||||
</nuclide>
|
||||
<nuclide name="Gd156" decay_modes="0" reactions="1">
|
||||
<reaction type="(n,gamma)" Q="0.0" target="Gd157" />
|
||||
</nuclide>
|
||||
<nuclide name="U234" decay_modes="0" reactions="1">
|
||||
<reaction type="fission" Q="191840000."/>
|
||||
<neutron_fission_yields>
|
||||
<energies>2.53000e-02</energies>
|
||||
<fission_yields energy="2.53000e-02">
|
||||
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
|
||||
<data>1.093250e-04 2.087260e-04 2.780820e-02 6.759540e-03 2.392300e-02 4.356330e-05</data>
|
||||
</fission_yields>
|
||||
</neutron_fission_yields>
|
||||
</nuclide>
|
||||
<nuclide name="U235" decay_modes="0" reactions="1">
|
||||
<reaction type="fission" Q="193410000."/>
|
||||
<neutron_fission_yields>
|
||||
<energies>2.53000e-02</energies>
|
||||
<fission_yields energy="2.53000e-02">
|
||||
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
|
||||
<data>6.142710e-5 1.483250e-04 0.0292737 0.002566345 0.0219242 4.9097e-6</data>
|
||||
</fission_yields>
|
||||
</neutron_fission_yields>
|
||||
</nuclide>
|
||||
<nuclide name="U238" decay_modes="0" reactions="1">
|
||||
<reaction type="fission" Q="197790000."/>
|
||||
<neutron_fission_yields>
|
||||
<energies>2.53000e-02</energies>
|
||||
<fission_yields energy="2.53000e-02">
|
||||
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
|
||||
<data>4.141120e-04 7.605360e-04 0.0135457 0.00026864 0.0024432 3.7100E-07</data>
|
||||
</fission_yields>
|
||||
</neutron_fission_yields>
|
||||
</nuclide>
|
||||
</depletion_chain>
|
||||
104
examples/python/pincell_depletion/restart_depletion.py
Normal file
104
examples/python/pincell_depletion/restart_depletion.py
Normal file
|
|
@ -0,0 +1,104 @@
|
|||
import openmc
|
||||
import openmc.deplete
|
||||
import numpy as np
|
||||
import matplotlib.pyplot as plt
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 100
|
||||
inactive = 10
|
||||
particles = 1000
|
||||
|
||||
# Depletion simulation parameters
|
||||
time_step = 1*24*60*60 # s
|
||||
final_time = 5*24*60*60 # s
|
||||
time_steps = np.full(final_time // time_step, time_step)
|
||||
|
||||
chain_file = './chain_simple.xml'
|
||||
power = 174 # W/cm, for 2D simulations only (use W for 3D)
|
||||
|
||||
###############################################################################
|
||||
# Load previous simulation results
|
||||
###############################################################################
|
||||
|
||||
# Load geometry from statepoint
|
||||
statepoint = 'statepoint.100.h5'
|
||||
with openmc.StatePoint(statepoint) as sp:
|
||||
geometry = sp.summary.geometry
|
||||
|
||||
# Load previous depletion results
|
||||
previous_results = openmc.deplete.ResultsList("depletion_results.h5")
|
||||
|
||||
###############################################################################
|
||||
# Transport calculation settings
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
entropy_mesh = openmc.RegularMesh()
|
||||
entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
|
||||
entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
|
||||
entropy_mesh.dimension = [10, 10, 1]
|
||||
settings_file.entropy_mesh = entropy_mesh
|
||||
|
||||
###############################################################################
|
||||
# Initialize and run depletion calculation
|
||||
###############################################################################
|
||||
|
||||
op = openmc.deplete.Operator(geometry, settings_file, chain_file,
|
||||
previous_results)
|
||||
|
||||
# Perform simulation using the predictor algorithm
|
||||
integrator = openmc.deplete.PredictorIntegrator(op, time_steps, power)
|
||||
integrator.integrate()
|
||||
|
||||
###############################################################################
|
||||
# Read depletion calculation results
|
||||
###############################################################################
|
||||
|
||||
# Open results file
|
||||
results = openmc.deplete.ResultsList.from_hdf5("depletion_results.h5")
|
||||
|
||||
# Obtain K_eff as a function of time
|
||||
time, keff = results.get_eigenvalue()
|
||||
|
||||
# Obtain U235 concentration as a function of time
|
||||
time, n_U235 = results.get_atoms('1', 'U235')
|
||||
|
||||
# Obtain Xe135 absorption as a function of time
|
||||
time, Xe_gam = results.get_reaction_rate('1', 'Xe135', '(n,gamma)')
|
||||
|
||||
###############################################################################
|
||||
# Generate plots
|
||||
###############################################################################
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), keff, label="K-effective")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("Keff")
|
||||
plt.show()
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), n_U235, label="U 235")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("n U5 (-)")
|
||||
plt.show()
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), Xe_gam, label="Xe135 absorption")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("RR (-)")
|
||||
plt.show()
|
||||
plt.close('all')
|
||||
175
examples/python/pincell_depletion/run_depletion.py
Normal file
175
examples/python/pincell_depletion/run_depletion.py
Normal file
|
|
@ -0,0 +1,175 @@
|
|||
import openmc
|
||||
import openmc.deplete
|
||||
import numpy as np
|
||||
import matplotlib.pyplot as plt
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 100
|
||||
inactive = 10
|
||||
particles = 1000
|
||||
|
||||
# Depletion simulation parameters
|
||||
time_step = 1*24*60*60 # s
|
||||
final_time = 5*24*60*60 # s
|
||||
time_steps = np.full(final_time // time_step, time_step)
|
||||
chain_file = './chain_simple.xml'
|
||||
power = 174 # W/cm, for 2D simulations only (use W for 3D)
|
||||
|
||||
###############################################################################
|
||||
# Define materials
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
|
||||
uo2.set_density('g/cm3', 10.29769)
|
||||
uo2.add_element('U', 1., enrichment=2.4)
|
||||
uo2.add_element('O', 2.)
|
||||
uo2.depletable = True
|
||||
|
||||
helium = openmc.Material(material_id=2, name='Helium for gap')
|
||||
helium.set_density('g/cm3', 0.001598)
|
||||
helium.add_element('He', 2.4044e-4)
|
||||
|
||||
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
|
||||
zircaloy.set_density('g/cm3', 6.55)
|
||||
zircaloy.add_element('Sn', 0.014 , 'wo')
|
||||
zircaloy.add_element('Fe', 0.00165, 'wo')
|
||||
zircaloy.add_element('Cr', 0.001 , 'wo')
|
||||
zircaloy.add_element('Zr', 0.98335, 'wo')
|
||||
|
||||
borated_water = openmc.Material(material_id=4, name='Borated water')
|
||||
borated_water.set_density('g/cm3', 0.740582)
|
||||
borated_water.add_element('B', 4.0e-5)
|
||||
borated_water.add_element('H', 5.0e-2)
|
||||
borated_water.add_element('O', 2.4e-2)
|
||||
borated_water.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
###############################################################################
|
||||
# Create geometry
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='Fuel OR')
|
||||
clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.40005, name='Clad IR')
|
||||
clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=0.45720, name='Clad OR')
|
||||
left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
|
||||
right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
|
||||
bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
|
||||
top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
|
||||
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
gap = openmc.Cell(cell_id=2, name='cell 2')
|
||||
clad = openmc.Cell(cell_id=3, name='cell 3')
|
||||
water = openmc.Cell(cell_id=4, name='cell 4')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
fuel.region = -fuel_or
|
||||
gap.region = +fuel_or & -clad_ir
|
||||
clad.region = +clad_ir & -clad_or
|
||||
water.region = +clad_or & +left & -right & +bottom & -top
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.fill = uo2
|
||||
gap.fill = helium
|
||||
clad.fill = zircaloy
|
||||
water.fill = borated_water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
root.add_cells([fuel, gap, clad, water])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe
|
||||
geometry = openmc.Geometry(root)
|
||||
|
||||
###############################################################################
|
||||
# Set volumes of depletable materials
|
||||
###############################################################################
|
||||
|
||||
# Compute cell areas
|
||||
area = {}
|
||||
area[fuel] = np.pi * fuel_or.coefficients['r'] ** 2
|
||||
|
||||
# Set materials volume for depletion. Set to an area for 2D simulations
|
||||
uo2.volume = area[fuel]
|
||||
|
||||
###############################################################################
|
||||
# Transport calculation settings
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
entropy_mesh = openmc.RegularMesh()
|
||||
entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
|
||||
entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
|
||||
entropy_mesh.dimension = [10, 10, 1]
|
||||
settings_file.entropy_mesh = entropy_mesh
|
||||
|
||||
###############################################################################
|
||||
# Initialize and run depletion calculation
|
||||
###############################################################################
|
||||
|
||||
op = openmc.deplete.Operator(geometry, settings_file, chain_file)
|
||||
|
||||
# Perform simulation using the predictor algorithm
|
||||
integrator = openmc.deplete.PredictorIntegrator(op, time_steps, power)
|
||||
integrator.integrate()
|
||||
|
||||
###############################################################################
|
||||
# Read depletion calculation results
|
||||
###############################################################################
|
||||
|
||||
# Open results file
|
||||
results = openmc.deplete.ResultsList.from_hdf5("depletion_results.h5")
|
||||
|
||||
# Obtain K_eff as a function of time
|
||||
time, keff = results.get_eigenvalue()
|
||||
|
||||
# Obtain U235 concentration as a function of time
|
||||
time, n_U235 = results.get_atoms('1', 'U235')
|
||||
|
||||
# Obtain Xe135 absorption as a function of time
|
||||
time, Xe_gam = results.get_reaction_rate('1', 'Xe135', '(n,gamma)')
|
||||
|
||||
###############################################################################
|
||||
# Generate plots
|
||||
###############################################################################
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), keff, label="K-effective")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("Keff")
|
||||
plt.show()
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), n_U235, label="U 235")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("n U5 (-)")
|
||||
plt.show()
|
||||
|
||||
plt.figure()
|
||||
plt.plot(time/(24*60*60), Xe_gam, label="Xe135 absorption")
|
||||
plt.xlabel("Time (days)")
|
||||
plt.ylabel("RR (-)")
|
||||
plt.show()
|
||||
plt.close('all')
|
||||
175
examples/python/pincell_multigroup/build-xml.py
Normal file
175
examples/python/pincell_multigroup/build-xml.py
Normal file
|
|
@ -0,0 +1,175 @@
|
|||
import numpy as np
|
||||
|
||||
import openmc
|
||||
import openmc.mgxs
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 100
|
||||
inactive = 10
|
||||
particles = 1000
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC mgxs.h5 file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate the energy group data
|
||||
groups = openmc.mgxs.EnergyGroups(group_edges=[
|
||||
1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6])
|
||||
|
||||
# Instantiate the 7-group (C5G7) cross section data
|
||||
uo2_xsdata = openmc.XSdata('UO2', groups)
|
||||
uo2_xsdata.order = 0
|
||||
uo2_xsdata.set_total(
|
||||
[0.1779492, 0.3298048, 0.4803882, 0.5543674, 0.3118013, 0.3951678,
|
||||
0.5644058])
|
||||
uo2_xsdata.set_absorption([8.0248E-03, 3.7174E-03, 2.6769E-02, 9.6236E-02,
|
||||
3.0020E-02, 1.1126E-01, 2.8278E-01])
|
||||
scatter_matrix = np.array(
|
||||
[[[0.1275370, 0.0423780, 0.0000094, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
|
||||
[0.0000000, 0.3244560, 0.0016314, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
|
||||
[0.0000000, 0.0000000, 0.4509400, 0.0026792, 0.0000000, 0.0000000, 0.0000000],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.4525650, 0.0055664, 0.0000000, 0.0000000],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0001253, 0.2714010, 0.0102550, 0.0000000],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0012968, 0.2658020, 0.0168090],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0085458, 0.2730800]]])
|
||||
scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
|
||||
uo2_xsdata.set_scatter_matrix(scatter_matrix)
|
||||
uo2_xsdata.set_fission([7.21206E-03, 8.19301E-04, 6.45320E-03,
|
||||
1.85648E-02, 1.78084E-02, 8.30348E-02,
|
||||
2.16004E-01])
|
||||
uo2_xsdata.set_nu_fission([2.005998E-02, 2.027303E-03, 1.570599E-02,
|
||||
4.518301E-02, 4.334208E-02, 2.020901E-01,
|
||||
5.257105E-01])
|
||||
uo2_xsdata.set_chi([5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07, 0.0000E+00,
|
||||
0.0000E+00, 0.0000E+00])
|
||||
|
||||
h2o_xsdata = openmc.XSdata('LWTR', groups)
|
||||
h2o_xsdata.order = 0
|
||||
h2o_xsdata.set_total([0.15920605, 0.412969593, 0.59030986, 0.58435,
|
||||
0.718, 1.2544497, 2.650379])
|
||||
h2o_xsdata.set_absorption([6.0105E-04, 1.5793E-05, 3.3716E-04,
|
||||
1.9406E-03, 5.7416E-03, 1.5001E-02,
|
||||
3.7239E-02])
|
||||
scatter_matrix = np.array(
|
||||
[[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0.0000000],
|
||||
[0.0000000, 0.2823340, 0.1299400, 0.0006234, 0.0000480, 0.0000074, 0.0000010],
|
||||
[0.0000000, 0.0000000, 0.3452560, 0.2245700, 0.0169990, 0.0026443, 0.0005034],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0910284, 0.4155100, 0.0637320, 0.0121390],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200],
|
||||
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]])
|
||||
scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
|
||||
h2o_xsdata.set_scatter_matrix(scatter_matrix)
|
||||
|
||||
mg_cross_sections_file = openmc.MGXSLibrary(groups)
|
||||
mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata])
|
||||
mg_cross_sections_file.export_to_hdf5()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Macroscopic Data
|
||||
uo2_data = openmc.Macroscopic('UO2')
|
||||
h2o_data = openmc.Macroscopic('LWTR')
|
||||
|
||||
# Instantiate some Materials and register the appropriate Macroscopic objects
|
||||
uo2 = openmc.Material(material_id=1, name='UO2 fuel')
|
||||
uo2.set_density('macro', 1.0)
|
||||
uo2.add_macroscopic(uo2_data)
|
||||
|
||||
water = openmc.Material(material_id=2, name='Water')
|
||||
water.set_density('macro', 1.0)
|
||||
water.add_macroscopic(h2o_data)
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([uo2, water])
|
||||
materials_file.cross_sections = "./mgxs.h5"
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.54, name='Fuel OR')
|
||||
left = openmc.XPlane(surface_id=4, x0=-0.63, name='left')
|
||||
right = openmc.XPlane(surface_id=5, x0=0.63, name='right')
|
||||
bottom = openmc.YPlane(surface_id=6, y0=-0.63, name='bottom')
|
||||
top = openmc.YPlane(surface_id=7, y0=0.63, name='top')
|
||||
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
moderator = openmc.Cell(cell_id=2, name='cell 2')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
fuel.region = -fuel_or
|
||||
moderator.region = +fuel_or & +left & -right & +bottom & -top
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.fill = uo2
|
||||
moderator.fill = water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
root.add_cells([fuel, moderator])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.energy_mode = "multi-group"
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.63, -0.63, -1, 0.63, 0.63, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh(mesh_id=1)
|
||||
mesh.dimension = [100, 100, 1]
|
||||
mesh.lower_left = [-0.63, -0.63, -1.e50]
|
||||
mesh.upper_right = [0.63, 0.63, 1.e50]
|
||||
|
||||
# Instantiate some tally Filters
|
||||
energy_filter = openmc.EnergyFilter([1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6,
|
||||
1.0e6, 20.0e6])
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1, name='tally 1')
|
||||
tally.filters = [energy_filter, mesh_filter]
|
||||
tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
|
||||
# Instantiate a Tallies collection, register all Tallies, and export to XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
82
examples/python/reflective/build-xml.py
Normal file
82
examples/python/reflective/build-xml.py
Normal file
|
|
@ -0,0 +1,82 @@
|
|||
import numpy as np
|
||||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 500
|
||||
inactive = 10
|
||||
particles = 10000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Material and register the Nuclide
|
||||
fuel = openmc.Material(material_id=1, name='fuel')
|
||||
fuel.set_density('g/cc', 4.5)
|
||||
fuel.add_nuclide('U235', 1.)
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([fuel])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate Surfaces
|
||||
surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
|
||||
surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
|
||||
surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
|
||||
surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
|
||||
surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
|
||||
surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
|
||||
|
||||
surf1.boundary_type = 'vacuum'
|
||||
surf2.boundary_type = 'vacuum'
|
||||
surf3.boundary_type = 'reflective'
|
||||
surf4.boundary_type = 'reflective'
|
||||
surf5.boundary_type = 'reflective'
|
||||
surf6.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cell
|
||||
cell = openmc.Cell(cell_id=1, name='cell 1')
|
||||
|
||||
# Use surface half-spaces to define region
|
||||
cell.region = +surf1 & -surf2 & +surf3 & -surf4 & +surf5 & -surf6
|
||||
|
||||
# Register Material with Cell
|
||||
cell.fill = fuel
|
||||
|
||||
# Instantiate Universes
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cell with Universe
|
||||
root.add_cell(cell)
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
uniform_dist = openmc.stats.Box(*cell.region.bounding_box,
|
||||
only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
Loading…
Add table
Add a link
Reference in a new issue