mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-27 13:45:36 -04:00
Rename main Python API classes to get rid of File.
This commit is contained in:
Paul Romano
parent
cccca4062a
commit
50a80693b5
37 changed files with 203 additions and 284 deletions
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@ -12,7 +12,7 @@ particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Nuclides
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@ -31,15 +31,15 @@ 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 MaterialsFile, register all Materials, and export to XML
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materials_file = openmc.MaterialsFile()
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# Instantiate a Materials collection, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate ZCylinder surfaces
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@ -74,22 +74,18 @@ cell1.fill = universe1
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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 and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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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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# 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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@ -103,7 +99,7 @@ settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC tallies.xml File
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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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@ -128,8 +124,8 @@ 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 TalliesFile, register all Tallies, and export to XML
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tallies_file = openmc.TalliesFile()
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# Instantiate a Tallies object, register all Tallies, and export to XML
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tallies_file = openmc.Tallies()
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tallies_file.add_tally(first_tally)
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tallies_file.add_tally(second_tally)
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tallies_file.add_tally(third_tally)
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@ -36,15 +36,15 @@ moderator.add_nuclide(h1, 2.)
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moderator.add_nuclide(o16, 1.)
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moderator.add_s_alpha_beta('HH2O', '71t')
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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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# Instantiate a Materials object, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate planar surfaces
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@ -97,14 +97,10 @@ outer_box.fill = moderator
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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 and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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@ -112,7 +108,7 @@ geometry_file.export_to_xml()
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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 = 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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@ -133,7 +129,7 @@ plot.width = [20, 20]
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plot.pixels = [200, 200]
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plot.color = 'cell'
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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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# Instantiate a Plots object, add Plot, and export to XML
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plot_file = openmc.Plots()
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plot_file.add_plot(plot)
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plot_file.export_to_xml()
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@ -35,15 +35,15 @@ 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_nuclide(fe56, 1.)
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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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# Instantiate a Materials object, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate Surfaces
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@ -105,22 +105,18 @@ 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 and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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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 = 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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@ -137,7 +133,7 @@ settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC plots.xml File
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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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@ -155,8 +151,8 @@ plot_yz.width = [8, 8]
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plot_yz.pixels = [400, 400]
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plot_yz.color = 'mat'
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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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# Instantiate a Plots object, add plots, and export to XML
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plot_file = openmc.Plots()
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plot_file.add_plot(plot_xy)
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plot_file.add_plot(plot_yz)
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plot_file.export_to_xml()
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@ -171,7 +167,7 @@ tally = openmc.Tally(tally_id=1)
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tally.filters = [openmc.Filter(type='distribcell', bins=[cell2.id])]
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tally.scores = ['total']
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# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
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tallies_file = openmc.TalliesFile()
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# Instantiate a Tallies object, register Tally/Mesh, and export to XML
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tallies_file = openmc.Tallies()
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tallies_file.add_tally(tally)
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tallies_file.export_to_xml()
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@ -11,7 +11,7 @@ particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Nuclides
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@ -30,15 +30,15 @@ moderator.add_nuclide(h1, 2.)
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moderator.add_nuclide(o16, 1.)
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moderator.add_s_alpha_beta('HH2O', '71t')
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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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# Instantiate a Materials object, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate Surfaces
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@ -116,22 +116,18 @@ lattice2.universes = [[univ4, univ4],
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cell1.fill = lattice2
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cell2.fill = lattice1
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# Instantiate a Geometry and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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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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# 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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@ -145,7 +141,7 @@ settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC plots.xml File
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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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@ -154,14 +150,14 @@ plot.width = [4, 4]
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plot.pixels = [400, 400]
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plot.color = 'mat'
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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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# Instantiate a Plots object, add Plot, and export to XML
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plot_file = openmc.Plots()
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plot_file.add_plot(plot)
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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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# Exporting to OpenMC tallies.xml file
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###############################################################################
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# Instantiate a tally mesh
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@ -180,8 +176,8 @@ tally = openmc.Tally(tally_id=1)
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tally.filters = [mesh_filter]
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tally.scores = ['total']
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# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
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tallies_file = openmc.TalliesFile()
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# Instantiate a Tallies object, register Tally/Mesh, and export to XML
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tallies_file = openmc.Tallies()
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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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@ -11,7 +11,7 @@ particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Nuclides
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@ -30,15 +30,15 @@ moderator.add_nuclide(h1, 2.)
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moderator.add_nuclide(o16, 1.)
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moderator.add_s_alpha_beta('HH2O', '71t')
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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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# Instantiate a Materials object, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate Surfaces
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@ -106,22 +106,18 @@ lattice.universes = [[univ1, univ2, univ1, univ2],
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# Fill Cell with the Lattice
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cell1.fill = lattice
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# Instantiate a Geometry and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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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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# 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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@ -137,7 +133,7 @@ settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC plots.xml File
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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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@ -146,14 +142,14 @@ plot.width = [4, 4]
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plot.pixels = [400, 400]
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plot.color = 'mat'
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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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# Instantiate a Plots object, add Plot, and export to XML
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plot_file = openmc.Plots()
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plot_file.add_plot(plot)
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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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# Exporting to OpenMC tallies.xml file
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###############################################################################
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# Instantiate a tally mesh
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@ -177,8 +173,8 @@ tally.filters = [mesh_filter]
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tally.scores = ['total']
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tally.triggers = [trigger]
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# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
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tallies_file = openmc.TalliesFile()
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# Instantiate a Tallies object, register Tally/Mesh, and export to XML
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tallies_file = openmc.Tallies()
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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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@ -11,7 +11,7 @@ particles = 1000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Nuclides
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@ -100,15 +100,15 @@ borated_water.add_nuclide(o16, 2.4672e-2)
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borated_water.add_nuclide(o17, 6.0099e-5)
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borated_water.add_s_alpha_beta('HH2O', '71t')
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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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# Instantiate a Materials object, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_materials([uo2, helium, zircaloy, borated_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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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate ZCylinder surfaces
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@ -149,22 +149,18 @@ 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, gap, clad, water])
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# Instantiate a Geometry and register the root Universe
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# Instantiate a Geometry and register the root Universe, and export to XML
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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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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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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()
|
||||
# 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
|
||||
|
|
@ -181,7 +177,7 @@ settings_file.export_to_xml()
|
|||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml File
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
|
|
@ -201,8 +197,8 @@ tally = openmc.Tally(tally_id=1, name='tally 1')
|
|||
tally.filters = [energy_filter, mesh_filter]
|
||||
tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
|
||||
# Instantiate a TalliesFile, register all Tallies, and export to XML
|
||||
tallies_file = openmc.TalliesFile()
|
||||
# Instantiate a Tallies object, register all Tallies, and export to XML
|
||||
tallies_file = openmc.Tallies()
|
||||
tallies_file.add_mesh(mesh)
|
||||
tallies_file.add_tally(tally)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
|
|||
|
|
@ -12,7 +12,7 @@ inactive = 10
|
|||
particles = 1000
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC mg_cross_sections.xml File
|
||||
# Exporting to OpenMC mg_cross_sections.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate the energy group data
|
||||
|
|
@ -59,13 +59,13 @@ scatter = [[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0
|
|||
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]]
|
||||
h2o_xsdata.scatter = np.array(scatter)
|
||||
|
||||
mg_cross_sections_file = openmc.MGXSLibraryFile(groups)
|
||||
mg_cross_sections_file = openmc.MGXSLibrary(groups)
|
||||
mg_cross_sections_file.add_xsdatas([uo2_xsdata,h2o_xsdata])
|
||||
mg_cross_sections_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml File
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Macroscopic Data
|
||||
|
|
@ -81,15 +81,15 @@ water = openmc.Material(material_id=2, name='Water')
|
|||
water.set_density('macro', 1.0)
|
||||
water.add_macroscopic(h2o_data)
|
||||
|
||||
# Instantiate a MaterialsFile, register all Materials, and export to XML
|
||||
materials_file = openmc.MaterialsFile()
|
||||
# Instantiate a Materials object, register all Materials, and export to XML
|
||||
materials_file = openmc.Materials()
|
||||
materials_file.default_xs = '300K'
|
||||
materials_file.add_materials([uo2, water])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml File
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
|
|
@ -122,22 +122,18 @@ root = openmc.Universe(universe_id=0, name='root universe')
|
|||
# Register Cells with Universe
|
||||
root.add_cells([fuel, moderator])
|
||||
|
||||
# Instantiate a Geometry and register the root Universe
|
||||
# Instantiate a Geometry and register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
# Instantiate a GeometryFile, register Geometry, and export to XML
|
||||
geometry_file = openmc.GeometryFile()
|
||||
geometry_file.geometry = geometry
|
||||
geometry_file.export_to_xml()
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml File
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.SettingsFile()
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.energy_mode = "multi-group"
|
||||
settings_file.cross_sections = "./mg_cross_sections.xml"
|
||||
settings_file.batches = batches
|
||||
|
|
@ -152,7 +148,7 @@ settings_file.source = openmc.source.Source(space=uniform_dist)
|
|||
settings_file.export_to_xml()
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml File
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
|
|
@ -177,8 +173,8 @@ tally.add_score('flux')
|
|||
tally.add_score('fission')
|
||||
tally.add_score('nu-fission')
|
||||
|
||||
# Instantiate a TalliesFile, register all Tallies, and export to XML
|
||||
tallies_file = openmc.TalliesFile()
|
||||
# Instantiate a Tallies object, register all Tallies, and export to XML
|
||||
tallies_file = openmc.Tallies()
|
||||
tallies_file.add_mesh(mesh)
|
||||
tallies_file.add_tally(tally)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
|
|||
|
|
@ -12,7 +12,7 @@ particles = 10000
|
|||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml File
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Nuclides
|
||||
|
|
@ -23,15 +23,15 @@ fuel = openmc.Material(material_id=1, name='fuel')
|
|||
fuel.set_density('g/cc', 4.5)
|
||||
fuel.add_nuclide(u235, 1.)
|
||||
|
||||
# Instantiate a MaterialsFile, register Material, and export to XML
|
||||
materials_file = openmc.MaterialsFile()
|
||||
# Instantiate a Materials object, register Material, and export to XML
|
||||
materials_file = openmc.Materials()
|
||||
materials_file.default_xs = '71c'
|
||||
materials_file.add_material(fuel)
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml File
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate Surfaces
|
||||
|
|
@ -64,22 +64,18 @@ root = openmc.Universe(universe_id=0, name='root universe')
|
|||
# Register Cell with Universe
|
||||
root.add_cell(cell)
|
||||
|
||||
# Instantiate a Geometry and register the root Universe
|
||||
# Instantiate a Geometry and register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
# Instantiate a GeometryFile, register Geometry, and export to XML
|
||||
geometry_file = openmc.GeometryFile()
|
||||
geometry_file.geometry = geometry
|
||||
geometry_file.export_to_xml()
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml File
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.SettingsFile()
|
||||
# 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
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue