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Adam Nelson
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examples/python/pincell_multigroup/build-xml.py
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examples/python/pincell_multigroup/build-xml.py
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import openmc
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import openmc.mgxs
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import numpy as np
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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 = 100
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inactive = 10
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particles = 1000
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###############################################################################
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# Exporting to OpenMC mg_cross_sections.xml File
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###############################################################################
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# Instantiate the energy group data
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groups = openmc.mgxs.EnergyGroups(group_edges=[1E-11, 0.0635E-6, 10.0E-6,
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1.0E-4, 1.0E-3, 0.5, 1.0, 20.0])
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# Instantiate the 7-group (C5G7) cross section data
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uo2_xsdata = openmc.Xsdata('UO2.300k', groups)
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uo2_xsdata.order = 0
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uo2_xsdata.total = np.array([0.1779492, 0.3298048, 0.4803882, 0.5543674,
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0.3118013, 0.3951678, 0.5644058])
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uo2_xsdata.absorption = np.array([8.0248E-03, 3.7174E-03, 2.6769E-02, 9.6236E-02,
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3.0020E-02, 1.1126E-01, 2.8278E-01])
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scatter = [[[0.1275370, 0.0423780, 0.0000094, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.3244560, 0.0016314, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.4509400, 0.0026792, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.4525650, 0.0055664, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.0001253, 0.2714010, 0.0102550, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0012968, 0.2658020, 0.0168090],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0085458, 0.2730800]]]
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uo2_xsdata.scatter = np.array(scatter[:][:])
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uo2_xsdata.fission = np.array([7.21206E-03, 8.19301E-04, 6.45320E-03,
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1.85648E-02, 1.78084E-02, 8.30348E-02,
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2.16004E-01])
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uo2_xsdata.nu_fission = np.array([2.005998E-02, 2.027303E-03, 1.570599E-02,
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4.518301E-02, 4.334208E-02, 2.020901E-01,
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5.257105E-01])
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uo2_xsdata.chi = np.array([5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07,
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0.0000E+00, 0.0000E+00, 0.0000E+00])
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h2o_xsdata = openmc.Xsdata('LWTR.300k', groups)
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h2o_xsdata.order = 0
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h2o_xsdata.total = np.array([0.15920605, 0.412969593, 0.59030986, 0.58435,
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0.718, 1.2544497, 2.650379])
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h2o_xsdata.absorption = np.array([6.0105E-04, 1.5793E-05, 3.3716E-04,
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1.9406E-03, 5.7416E-03, 1.5001E-02,
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3.7239E-02])
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scatter = [[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0.0000000],
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[0.0000000, 0.2823340, 0.1299400, 0.0006234, 0.0000480, 0.0000074, 0.0000010],
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[0.0000000, 0.0000000, 0.3452560, 0.2245700, 0.0169990, 0.0026443, 0.0005034],
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[0.0000000, 0.0000000, 0.0000000, 0.0910284, 0.4155100, 0.0637320, 0.0121390],
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[0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]]
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h2o_xsdata.scatter = np.array(scatter[:][:])
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mg_cross_sections_file = openmc.MGXSLibraryFile(groups)
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mg_cross_sections_file.add_xsdatas([uo2_xsdata,h2o_xsdata])
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mg_cross_sections_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC materials.xml File
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###############################################################################
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# Instantiate some Macroscopic Data
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uo2_data = openmc.Macroscopic('UO2', '300k')
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h2o_data = openmc.Macroscopic('LWTR', '300k')
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# Instantiate some Materials and register the appropriate Nuclides
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uo2 = openmc.Material(material_id=1, name='UO2 fuel')
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uo2.set_density('macro', 1.0)
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uo2.add_macroscopic(uo2_data)
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water = openmc.Material(material_id=2, name='Water')
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water.set_density('macro', 1.0)
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water.add_macroscopic(h2o_data)
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# Instantiate a MaterialsFile, register all Materials, and export to XML
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materials_file = openmc.MaterialsFile()
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materials_file.default_xs = '300k'
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materials_file.add_materials([uo2, water])
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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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fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.54, name='Fuel OR')
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left = openmc.XPlane(surface_id=4, x0=-0.63, name='left')
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right = openmc.XPlane(surface_id=5, x0=0.63, name='right')
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bottom = openmc.YPlane(surface_id=6, y0=-0.63, name='bottom')
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top = openmc.YPlane(surface_id=7, y0=0.63, name='top')
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left.boundary_type = 'reflective'
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right.boundary_type = 'reflective'
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top.boundary_type = 'reflective'
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bottom.boundary_type = 'reflective'
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# Instantiate Cells
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fuel = openmc.Cell(cell_id=1, name='cell 1')
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moderator = openmc.Cell(cell_id=2, name='cell 2')
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# Use surface half-spaces to define regions
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fuel.region = -fuel_or
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moderator.region = +fuel_or & +left & -right & +bottom & -top
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# Register Materials with Cells
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fuel.fill = uo2
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moderator.fill = water
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# Instantiate Universe
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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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root.add_cells([fuel, moderator])
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.root_universe = root
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# Instantiate a GeometryFile, register Geometry, and export to XML
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geometry_file = openmc.GeometryFile()
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geometry_file.geometry = geometry
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geometry_file.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 SettingsFile, set all runtime parameters, and export to XML
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settings_file = openmc.SettingsFile()
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settings_file.energy_mode = "multi-group"
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settings_file.cross_sections = "./mg_cross_sections.xml"
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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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settings_file.set_source_space('box', [-0.63, -0.63, -1, \
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0.63, 0.63, 1])
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settings_file.entropy_lower_left = [-0.54, -0.54, -1.e50]
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settings_file.entropy_upper_right = [0.54, 0.54, 1.e50]
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settings_file.entropy_dimension = [10, 10, 1]
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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 a tally mesh
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mesh = openmc.Mesh(mesh_id=1)
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mesh.type = 'regular'
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mesh.dimension = [100, 100, 1]
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mesh.lower_left = [-0.63, -0.63, -1.e50]
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mesh.upper_right = [0.63, 0.63, 1.e50]
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# Instantiate some tally Filters
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# energy_filter = openmc.Filter(type='energy',
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# bins=[1E-11, 0.0635E-6, 10.0E-6, 1.0E-4, 1.0E-3,
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# 0.5, 1.0, 20.0])
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energy_filter = openmc.Filter(type='energy')
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mesh_filter = openmc.Filter()
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mesh_filter.mesh = mesh
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# Instantiate the Tally
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tally = openmc.Tally(tally_id=1, name='tally 1')
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tally.add_filter(energy_filter)
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tally.add_filter(mesh_filter)
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tally.add_score('flux')
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tally.add_score('fission')
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tally.add_score('nu-fission')
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# Instantiate a TalliesFile, register all Tallies, and export to XML
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tallies_file = openmc.TalliesFile()
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tallies_file.add_mesh(mesh)
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tallies_file.add_tally(tally)
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tallies_file.export_to_xml()
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