mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-28 22:26:08 -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
6
docs/source/_templates/myfunction.rst
Normal file
6
docs/source/_templates/myfunction.rst
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@ -0,0 +1,6 @@
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{{ fullname }}
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{{ underline }}
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.. currentmodule:: {{ module }}
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.. autofunction:: {{ objname }}
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@ -29,7 +29,7 @@ Classes
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:template: myclass.rst
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openmc.XSdata
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openmc.MGXSLibraryFile
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openmc.MGXSLibrary
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Functions
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+++++++++
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@ -50,7 +50,7 @@ Simulation Settings
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openmc.Source
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openmc.ResonanceScattering
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openmc.SettingsFile
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openmc.Settings
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Material Specification
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----------------------
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@ -64,7 +64,7 @@ Material Specification
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openmc.Element
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openmc.Macroscopic
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openmc.Material
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openmc.MaterialsFile
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openmc.Materials
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Building geometry
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-----------------
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@ -96,7 +96,6 @@ Building geometry
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openmc.RectLattice
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openmc.HexLattice
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openmc.Geometry
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openmc.GeometryFile
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Many of the above classes are derived from several abstract classes:
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@ -121,7 +120,7 @@ Constructing Tallies
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openmc.Mesh
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openmc.Trigger
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openmc.Tally
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openmc.TalliesFile
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openmc.Tallies
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Coarse Mesh Finite Difference Acceleration
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------------------------------------------
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@ -132,7 +131,7 @@ Coarse Mesh Finite Difference Acceleration
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:template: myclass.rst
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openmc.CMFDMesh
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openmc.CMFDFile
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openmc.CMFD
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Plotting
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--------
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@ -143,7 +142,7 @@ Plotting
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:template: myclass.rst
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openmc.Plot
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openmc.PlotsFile
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openmc.Plots
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Running OpenMC
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--------------
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@ -151,9 +150,10 @@ Running OpenMC
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.. autosummary::
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:toctree: generated
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:nosignatures:
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:template: myclass.rst
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:template: myfunction.rst
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openmc.Executor
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openmc.run
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openmc.plot_geometry
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Post-processing
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---------------
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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
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
|
|
@ -177,8 +173,8 @@ tally.filters = [mesh_filter]
|
|||
tally.scores = ['total']
|
||||
tally.triggers = [trigger]
|
||||
|
||||
# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
|
||||
tallies_file = openmc.TalliesFile()
|
||||
# Instantiate a Tallies object, register Tally/Mesh, and export to XML
|
||||
tallies_file = openmc.Tallies()
|
||||
tallies_file.add_mesh(mesh)
|
||||
tallies_file.add_tally(tally)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
|
|||
|
|
@ -11,7 +11,7 @@ particles = 1000
|
|||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml File
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Nuclides
|
||||
|
|
@ -100,15 +100,15 @@ borated_water.add_nuclide(o16, 2.4672e-2)
|
|||
borated_water.add_nuclide(o17, 6.0099e-5)
|
||||
borated_water.add_s_alpha_beta('HH2O', '71t')
|
||||
|
||||
# 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 = '71c'
|
||||
materials_file.add_materials([uo2, helium, zircaloy, borated_water])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml File
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
|
|
@ -149,22 +149,18 @@ root = openmc.Universe(universe_id=0, name='root universe')
|
|||
# Register Cells with Universe
|
||||
root.add_cells([fuel, gap, clad, water])
|
||||
|
||||
# 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
|
||||
|
|
@ -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
|
||||
|
|
|
|||
|
|
@ -187,7 +187,7 @@ class CMFDMesh(object):
|
|||
return element
|
||||
|
||||
|
||||
class CMFDFile(object):
|
||||
class CMFD(object):
|
||||
"""Parameters that control the use of coarse-mesh finite difference acceleration
|
||||
in OpenMC. This corresponds directly to the cmfd.xml input file.
|
||||
|
||||
|
|
|
|||
|
|
@ -27,7 +27,7 @@ def _run(command, output, cwd):
|
|||
return p.returncode
|
||||
|
||||
|
||||
def plot(output=True, openmc_exec='openmc', cwd='.'):
|
||||
def plot_geometry(output=True, openmc_exec='openmc', cwd='.'):
|
||||
"""Run OpenMC in plotting mode
|
||||
|
||||
Parameters
|
||||
|
|
|
|||
|
|
@ -23,7 +23,6 @@ class Geometry(object):
|
|||
"""
|
||||
|
||||
def __init__(self):
|
||||
# Initialize Geometry class attributes
|
||||
self._root_universe = None
|
||||
self._offsets = {}
|
||||
|
||||
|
|
@ -42,6 +41,27 @@ class Geometry(object):
|
|||
|
||||
self._root_universe = root_universe
|
||||
|
||||
def export_to_xml(self):
|
||||
"""Create a geometry.xml file that can be used for a simulation.
|
||||
|
||||
"""
|
||||
|
||||
# Clear OpenMC written IDs used to optimize XML generation
|
||||
openmc.universe.WRITTEN_IDS = {}
|
||||
|
||||
# Create XML representation
|
||||
geometry_file = ET.Element("geometry")
|
||||
self.root_universe.create_xml_subelement(geometry_file)
|
||||
|
||||
# Clean the indentation in the file to be user-readable
|
||||
sort_xml_elements(geometry_file)
|
||||
clean_xml_indentation(geometry_file)
|
||||
|
||||
# Write the XML Tree to the geometry.xml file
|
||||
tree = ET.ElementTree(geometry_file)
|
||||
tree.write("geometry.xml", xml_declaration=True, encoding='utf-8',
|
||||
method="xml")
|
||||
|
||||
def get_cell_instance(self, path):
|
||||
"""Return the instance number for the final cell in a geometry path.
|
||||
|
||||
|
|
@ -436,52 +456,3 @@ class Geometry(object):
|
|||
lattices = list(lattices)
|
||||
lattices.sort(key=lambda x: x.id)
|
||||
return lattices
|
||||
|
||||
|
||||
class GeometryFile(object):
|
||||
"""Geometry file used for an OpenMC simulation. Corresponds directly to the
|
||||
geometry.xml input file.
|
||||
|
||||
Attributes
|
||||
----------
|
||||
geometry : openmc.Geometry
|
||||
The geometry to be used
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self):
|
||||
# Initialize GeometryFile class attributes
|
||||
self._geometry = None
|
||||
self._geometry_file = ET.Element("geometry")
|
||||
|
||||
@property
|
||||
def geometry(self):
|
||||
return self._geometry
|
||||
|
||||
@geometry.setter
|
||||
def geometry(self, geometry):
|
||||
check_type('the geometry', geometry, Geometry)
|
||||
self._geometry = geometry
|
||||
|
||||
def export_to_xml(self):
|
||||
"""Create a geometry.xml file that can be used for a simulation.
|
||||
|
||||
"""
|
||||
|
||||
# Clear OpenMC written IDs used to optimize XML generation
|
||||
openmc.universe.WRITTEN_IDS = {}
|
||||
|
||||
# Reset xml element tree
|
||||
self._geometry_file.clear()
|
||||
|
||||
root_universe = self.geometry.root_universe
|
||||
root_universe.create_xml_subelement(self._geometry_file)
|
||||
|
||||
# Clean the indentation in the file to be user-readable
|
||||
sort_xml_elements(self._geometry_file)
|
||||
clean_xml_indentation(self._geometry_file)
|
||||
|
||||
# Write the XML Tree to the geometry.xml file
|
||||
tree = ET.ElementTree(self._geometry_file)
|
||||
tree.write("geometry.xml", xml_declaration=True,
|
||||
encoding='utf-8', method="xml")
|
||||
|
|
|
|||
|
|
@ -642,8 +642,8 @@ class Material(object):
|
|||
return element
|
||||
|
||||
|
||||
class MaterialsFile(object):
|
||||
"""Materials file used for an OpenMC simulation. Corresponds directly to the
|
||||
class Materials(object):
|
||||
"""Materials used for an OpenMC simulation. Corresponds directly to the
|
||||
materials.xml input file.
|
||||
|
||||
Attributes
|
||||
|
|
@ -655,7 +655,6 @@ class MaterialsFile(object):
|
|||
"""
|
||||
|
||||
def __init__(self):
|
||||
# Initialize MaterialsFile class attributes
|
||||
self._materials = []
|
||||
self._default_xs = None
|
||||
self._materials_file = ET.Element("materials")
|
||||
|
|
@ -681,7 +680,7 @@ class MaterialsFile(object):
|
|||
|
||||
if not isinstance(material, Material):
|
||||
msg = 'Unable to add a non-Material "{0}" to the ' \
|
||||
'MaterialsFile'.format(material)
|
||||
'Materials instance'.format(material)
|
||||
raise ValueError(msg)
|
||||
|
||||
self._materials.append(material)
|
||||
|
|
@ -716,7 +715,7 @@ class MaterialsFile(object):
|
|||
|
||||
if not isinstance(material, Material):
|
||||
msg = 'Unable to remove a non-Material "{0}" from the ' \
|
||||
'MaterialsFile'.format(material)
|
||||
'Materials instance'.format(material)
|
||||
raise ValueError(msg)
|
||||
|
||||
self._materials.remove(material)
|
||||
|
|
|
|||
|
|
@ -354,8 +354,8 @@ class Library(object):
|
|||
|
||||
Parameters
|
||||
----------
|
||||
tallies_file : openmc.TalliesFile
|
||||
A TalliesFile object to add each MGXS' tallies to generate a
|
||||
tallies_file : openmc.Tallies
|
||||
A Tallies object to add each MGXS' tallies to generate a
|
||||
"tallies.xml" input file for OpenMC
|
||||
merge : bool
|
||||
Indicate whether tallies should be merged when possible. Defaults
|
||||
|
|
@ -363,7 +363,7 @@ class Library(object):
|
|||
|
||||
"""
|
||||
|
||||
cv.check_type('tallies_file', tallies_file, openmc.TalliesFile)
|
||||
cv.check_type('tallies_file', tallies_file, openmc.Tallies)
|
||||
|
||||
# Add tallies from each MGXS for each domain and mgxs type
|
||||
for domain in self.domains:
|
||||
|
|
|
|||
|
|
@ -647,7 +647,7 @@ class XSdata(object):
|
|||
return element
|
||||
|
||||
|
||||
class MGXSLibraryFile(object):
|
||||
class MGXSLibrary(object):
|
||||
"""Multi-Group Cross Sections file used for an OpenMC simulation.
|
||||
Corresponds directly to the MG version of the cross_sections.xml input file.
|
||||
|
||||
|
|
@ -662,7 +662,6 @@ class MGXSLibraryFile(object):
|
|||
"""
|
||||
|
||||
def __init__(self, energy_groups):
|
||||
# Initialize MGXSLibraryFile class attributes
|
||||
self._xsdatas = []
|
||||
self._energy_groups = energy_groups
|
||||
self._inverse_velocities = None
|
||||
|
|
@ -701,12 +700,12 @@ class MGXSLibraryFile(object):
|
|||
# Check the type
|
||||
if not isinstance(xsdata, XSdata):
|
||||
msg = 'Unable to add a non-XSdata "{0}" to the ' \
|
||||
'MGXSLibraryFile'.format(xsdata)
|
||||
'MGXSLibrary instance'.format(xsdata)
|
||||
raise ValueError(msg)
|
||||
|
||||
# Make sure energy groups match.
|
||||
if xsdata.energy_groups != self._energy_groups:
|
||||
msg = 'Energy groups of XSdata do not match that of MGXSLibraryFile!'
|
||||
msg = 'Energy groups of XSdata do not match that of MGXSLibrary!'
|
||||
raise ValueError(msg)
|
||||
|
||||
self._xsdatas.append(xsdata)
|
||||
|
|
@ -741,7 +740,7 @@ class MGXSLibraryFile(object):
|
|||
|
||||
if not isinstance(xsdata, XSdata):
|
||||
msg = 'Unable to remove a non-XSdata "{0}" from the ' \
|
||||
'XSdatasFile'.format(xsdata)
|
||||
'MGXSLibrary instance'.format(xsdata)
|
||||
raise ValueError(msg)
|
||||
|
||||
self._xsdatas.remove(xsdata)
|
||||
|
|
|
|||
|
|
@ -401,14 +401,13 @@ class Plot(object):
|
|||
return element
|
||||
|
||||
|
||||
class PlotsFile(object):
|
||||
class Plots(object):
|
||||
"""Plots file used for an OpenMC simulation. Corresponds directly to the
|
||||
plots.xml input file.
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self):
|
||||
# Initialize PlotsFile class attributes
|
||||
self._plots = []
|
||||
self._plots_file = ET.Element("plots")
|
||||
|
||||
|
|
@ -423,7 +422,7 @@ class PlotsFile(object):
|
|||
"""
|
||||
|
||||
if not isinstance(plot, Plot):
|
||||
msg = 'Unable to add a non-Plot "{0}" to the PlotsFile'.format(plot)
|
||||
msg = 'Unable to add a non-Plot "{0}" to the Plots instance'.format(plot)
|
||||
raise ValueError(msg)
|
||||
|
||||
self._plots.append(plot)
|
||||
|
|
|
|||
|
|
@ -16,7 +16,7 @@ if sys.version_info[0] >= 3:
|
|||
basestring = str
|
||||
|
||||
|
||||
class SettingsFile(object):
|
||||
class Settings(object):
|
||||
"""Settings file used for an OpenMC simulation. Corresponds directly to the
|
||||
settings.xml input file.
|
||||
|
||||
|
|
|
|||
|
|
@ -3419,14 +3419,13 @@ class Tally(object):
|
|||
return new_tally
|
||||
|
||||
|
||||
class TalliesFile(object):
|
||||
class Tallies(object):
|
||||
"""Tallies file used for an OpenMC simulation. Corresponds directly to the
|
||||
tallies.xml input file.
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self):
|
||||
# Initialize TalliesFile class attributes
|
||||
self._tallies = []
|
||||
self._meshes = []
|
||||
self._tallies_file = ET.Element("tallies")
|
||||
|
|
@ -3453,7 +3452,7 @@ class TalliesFile(object):
|
|||
"""
|
||||
|
||||
if not isinstance(tally, Tally):
|
||||
msg = 'Unable to add a non-Tally "{0}" to the TalliesFile'.format(tally)
|
||||
msg = 'Unable to add a non-Tally "{0}" to the Tallies instance'.format(tally)
|
||||
raise ValueError(msg)
|
||||
|
||||
if merge:
|
||||
|
|
@ -3524,7 +3523,7 @@ class TalliesFile(object):
|
|||
"""
|
||||
|
||||
if not isinstance(mesh, Mesh):
|
||||
msg = 'Unable to add a non-Mesh "{0}" to the TalliesFile'.format(mesh)
|
||||
msg = 'Unable to add a non-Mesh "{0}" to the Tallies instance'.format(mesh)
|
||||
raise ValueError(msg)
|
||||
|
||||
self._meshes.append(mesh)
|
||||
|
|
|
|||
|
|
@ -5,9 +5,9 @@ from openmc.stats import Box
|
|||
|
||||
class InputSet(object):
|
||||
def __init__(self):
|
||||
self.settings = openmc.SettingsFile()
|
||||
self.materials = openmc.MaterialsFile()
|
||||
self.geometry = openmc.GeometryFile()
|
||||
self.settings = openmc.Settings()
|
||||
self.materials = openmc.Materials()
|
||||
self.geometry = openmc.Geometry()
|
||||
self.tallies = None
|
||||
self.plots = None
|
||||
|
||||
|
|
@ -550,11 +550,8 @@ class InputSet(object):
|
|||
|
||||
root.add_cells((c1, c2, c3, c4, c5, c6, c7, c8, c9, c10, c11, c12))
|
||||
|
||||
# Define the geometry file.
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
self.geometry.geometry = geometry
|
||||
# Assign root universe to geometry
|
||||
self.geometry.root_universe = root
|
||||
|
||||
def build_default_settings(self):
|
||||
self.settings.batches = 10
|
||||
|
|
@ -630,12 +627,8 @@ class MGInputSet(InputSet):
|
|||
|
||||
root.add_cells((c1,c2,c3))
|
||||
|
||||
# Define the geometry file.
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
self.geometry.geometry = geometry
|
||||
|
||||
# Assign root universe to geometry
|
||||
self.geometry.root_universe = root
|
||||
|
||||
def build_default_settings(self):
|
||||
self.settings.batches = 10
|
||||
|
|
@ -656,8 +649,3 @@ class MGInputSet(InputSet):
|
|||
plot.color = 'mat'
|
||||
|
||||
self.plots.add_plot(plot)
|
||||
|
||||
|
||||
|
||||
|
||||
|
||||
|
|
|
|||
|
|
@ -7,8 +7,6 @@ import hashlib
|
|||
sys.path.insert(0, os.pardir)
|
||||
from testing_harness import PyAPITestHarness
|
||||
import openmc
|
||||
from openmc.source import Source
|
||||
from openmc.stats import Box
|
||||
|
||||
|
||||
class AsymmetricLatticeTestHarness(PyAPITestHarness):
|
||||
|
|
@ -20,7 +18,7 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
|
|||
self._input_set.build_default_materials_and_geometry()
|
||||
|
||||
# Extract universes encapsulating fuel and water assemblies
|
||||
geometry = self._input_set.geometry.geometry
|
||||
geometry = self._input_set.geometry
|
||||
water = geometry.get_universes_by_name('water assembly (hot)')[0]
|
||||
fuel = geometry.get_universes_by_name('fuel assembly (hot)')[0]
|
||||
|
||||
|
|
@ -49,7 +47,7 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
|
|||
root_univ.add_cell(root_cell)
|
||||
|
||||
# Over-ride geometry in the input set with this 3x3 lattice
|
||||
self._input_set.geometry.geometry.root_universe = root_univ
|
||||
self._input_set.geometry.root_universe = root_univ
|
||||
|
||||
# Initialize a "distribcell" filter for the fuel pin cell
|
||||
distrib_filter = openmc.Filter(type='distribcell', bins=[27])
|
||||
|
|
@ -60,7 +58,7 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
|
|||
tally.add_score('nu-fission')
|
||||
|
||||
# Initialize the tallies file
|
||||
tallies_file = openmc.TalliesFile()
|
||||
tallies_file = openmc.Tallies()
|
||||
tallies_file.add_tally(tally)
|
||||
|
||||
# Assign the tallies file to the input set
|
||||
|
|
@ -70,7 +68,7 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
|
|||
self._input_set.build_default_settings()
|
||||
|
||||
# Specify summary output and correct source sampling box
|
||||
source = Source(space=Box([-32, -32, 0], [32, 32, 32]))
|
||||
source = openmc.Source(space=openmc.stats.Box([-32, -32, 0], [32, 32, 32]))
|
||||
source.space.only_fissionable = True
|
||||
self._input_set.settings.source = source
|
||||
self._input_set.settings.output = {'summary': True}
|
||||
|
|
|
|||
|
|
@ -28,7 +28,7 @@ class DistribmatTestHarness(PyAPITestHarness):
|
|||
light_fuel.set_density('g/cc', 2.0)
|
||||
light_fuel.add_nuclide('U-235', 1.0)
|
||||
|
||||
mats_file = openmc.MaterialsFile()
|
||||
mats_file = openmc.Materials()
|
||||
mats_file.default_xs = '71c'
|
||||
mats_file.add_materials([moderator, dense_fuel, light_fuel])
|
||||
mats_file.export_to_xml()
|
||||
|
|
@ -74,16 +74,14 @@ class DistribmatTestHarness(PyAPITestHarness):
|
|||
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root_univ
|
||||
geo_file = openmc.GeometryFile()
|
||||
geo_file.geometry = geometry
|
||||
geo_file.export_to_xml()
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
####################
|
||||
# Settings
|
||||
####################
|
||||
|
||||
sets_file = openmc.SettingsFile()
|
||||
sets_file = openmc.Settings()
|
||||
sets_file.batches = 5
|
||||
sets_file.inactive = 0
|
||||
sets_file.particles = 1000
|
||||
|
|
@ -96,7 +94,7 @@ class DistribmatTestHarness(PyAPITestHarness):
|
|||
# Plots
|
||||
####################
|
||||
|
||||
plots_file = openmc.PlotsFile()
|
||||
plots_file = openmc.Plots()
|
||||
|
||||
plot = openmc.Plot(plot_id=1)
|
||||
plot.basis = 'xy'
|
||||
|
|
|
|||
|
|
@ -68,7 +68,7 @@ class MGNuclideInputSet(MGInputSet):
|
|||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
self.geometry.geometry = geometry
|
||||
self.geometry = geometry
|
||||
|
||||
class MGMaxOrderTestHarness(PyAPITestHarness):
|
||||
def __init__(self, statepoint_name, tallies_present, mg=False):
|
||||
|
|
|
|||
|
|
@ -67,7 +67,7 @@ class MGNuclideInputSet(MGInputSet):
|
|||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
|
||||
self.geometry.geometry = geometry
|
||||
self.geometry = geometry
|
||||
|
||||
class MGNuclideTestHarness(PyAPITestHarness):
|
||||
def __init__(self, statepoint_name, tallies_present, mg=False):
|
||||
|
|
|
|||
|
|
@ -41,7 +41,7 @@ class MGTalliesTestHarness(PyAPITestHarness):
|
|||
tally2.add_score('scatter')
|
||||
tally2.add_score('nu-scatter')
|
||||
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self._input_set.tallies.add_mesh(mesh)
|
||||
self._input_set.tallies.add_tally(tally1)
|
||||
self._input_set.tallies.add_tally(tally2)
|
||||
|
|
|
|||
|
|
@ -23,7 +23,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625e-6, 20.])
|
||||
|
||||
# Initialize MGXS Library for a few cross section types
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry.geometry)
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
|
||||
self.mgxs_lib.by_nuclide = False
|
||||
self.mgxs_lib.mgxs_types = ['transport', 'nu-fission',
|
||||
'nu-scatter matrix', 'chi']
|
||||
|
|
@ -32,7 +32,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib.build_library()
|
||||
|
||||
# Initialize a tallies file
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
|
||||
self._input_set.tallies.export_to_xml()
|
||||
|
||||
|
|
|
|||
|
|
@ -24,7 +24,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
|
||||
# Initialize MGXS Library for a few cross section types
|
||||
# for one material-filled cell in the geometry
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry.geometry)
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
|
||||
self.mgxs_lib.by_nuclide = False
|
||||
self.mgxs_lib.mgxs_types = ['transport', 'nu-fission',
|
||||
'nu-scatter matrix', 'chi']
|
||||
|
|
@ -35,7 +35,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib.build_library()
|
||||
|
||||
# Initialize a tallies file
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
|
||||
self._input_set.tallies.export_to_xml()
|
||||
|
||||
|
|
|
|||
|
|
@ -24,7 +24,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625e-6, 20.])
|
||||
|
||||
# Initialize MGXS Library for a few cross section types
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry.geometry)
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
|
||||
self.mgxs_lib.by_nuclide = False
|
||||
self.mgxs_lib.mgxs_types = ['transport', 'nu-fission',
|
||||
'nu-scatter matrix', 'chi']
|
||||
|
|
@ -33,7 +33,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib.build_library()
|
||||
|
||||
# Initialize a tallies file
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
|
||||
self._input_set.tallies.export_to_xml()
|
||||
|
||||
|
|
@ -51,7 +51,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
|
||||
# Load the MGXS library from the statepoint
|
||||
self.mgxs_lib.load_from_statepoint(sp)
|
||||
|
||||
|
||||
# Export the MGXS Library to an HDF5 file
|
||||
self.mgxs_lib.build_hdf5_store(directory='.')
|
||||
|
||||
|
|
@ -67,7 +67,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
outstr += str(f[key][...]) + '\n'
|
||||
key = 'material/{0}/{1}/std. dev.'.format(domain.id, mgxs_type)
|
||||
outstr += str(f[key][...]) + '\n'
|
||||
|
||||
|
||||
# Close the MGXS HDF5 file
|
||||
f.close()
|
||||
|
||||
|
|
|
|||
|
|
@ -23,7 +23,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625e-6, 20.])
|
||||
|
||||
# Initialize MGXS Library for a few cross section types
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry.geometry)
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
|
||||
self.mgxs_lib.by_nuclide = False
|
||||
self.mgxs_lib.mgxs_types = ['transport', 'nu-fission',
|
||||
'nu-scatter matrix', 'chi']
|
||||
|
|
@ -32,7 +32,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib.build_library()
|
||||
|
||||
# Initialize a tallies file
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
|
||||
self._input_set.tallies.export_to_xml()
|
||||
|
||||
|
|
|
|||
|
|
@ -23,7 +23,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625e-6, 20.])
|
||||
|
||||
# Initialize MGXS Library for a few cross section types
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry.geometry)
|
||||
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
|
||||
self.mgxs_lib.by_nuclide = True
|
||||
self.mgxs_lib.mgxs_types = ['transport', 'nu-fission',
|
||||
'nu-scatter matrix', 'chi']
|
||||
|
|
@ -32,7 +32,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib.build_library()
|
||||
|
||||
# Initialize a tallies file
|
||||
self._input_set.tallies = openmc.TalliesFile()
|
||||
self._input_set.tallies = openmc.Tallies()
|
||||
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
|
||||
self._input_set.tallies.export_to_xml()
|
||||
|
||||
|
|
|
|||
|
|
@ -19,7 +19,7 @@ class PlotTestHarness(TestHarness):
|
|||
self._plot_names = plot_names
|
||||
|
||||
def _run_openmc(self):
|
||||
returncode = openmc.plot(openmc_exec=self._opts.exe)
|
||||
returncode = openmc.plot_geometry(openmc_exec=self._opts.exe)
|
||||
assert returncode == 0, 'OpenMC did not exit successfully.'
|
||||
|
||||
def _test_output_created(self):
|
||||
|
|
|
|||
|
|
@ -17,7 +17,7 @@ class ResonanceScatteringTestHarness(PyAPITestHarness):
|
|||
mat.add_nuclide('Pu-239', 0.02)
|
||||
mat.add_nuclide('H-1', 20.0)
|
||||
|
||||
mats_file = openmc.MaterialsFile()
|
||||
mats_file = openmc.Materials()
|
||||
mats_file.default_xs = '71c'
|
||||
mats_file.add_material(mat)
|
||||
mats_file.export_to_xml()
|
||||
|
|
@ -35,9 +35,7 @@ class ResonanceScatteringTestHarness(PyAPITestHarness):
|
|||
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root_univ
|
||||
geo_file = openmc.GeometryFile()
|
||||
geo_file.geometry = geometry
|
||||
geo_file.export_to_xml()
|
||||
geometry.export_to_xml()
|
||||
|
||||
# Settings
|
||||
nuclide = openmc.Nuclide('U-238', '71c')
|
||||
|
|
@ -67,7 +65,7 @@ class ResonanceScatteringTestHarness(PyAPITestHarness):
|
|||
res_scatt_ares.E_min = 1e-6
|
||||
res_scatt_ares.E_max = 210e-6
|
||||
|
||||
sets_file = openmc.SettingsFile()
|
||||
sets_file = openmc.Settings()
|
||||
sets_file.batches = 10
|
||||
sets_file.inactive = 5
|
||||
sets_file.particles = 1000
|
||||
|
|
|
|||
|
|
@ -9,8 +9,6 @@ import numpy as np
|
|||
sys.path.insert(0, os.pardir)
|
||||
from testing_harness import PyAPITestHarness
|
||||
import openmc
|
||||
import openmc.stats
|
||||
from openmc.source import Source
|
||||
|
||||
|
||||
class SourceTestHarness(PyAPITestHarness):
|
||||
|
|
@ -18,7 +16,7 @@ class SourceTestHarness(PyAPITestHarness):
|
|||
mat1 = openmc.Material(material_id=1)
|
||||
mat1.set_density('g/cm3', 4.5)
|
||||
mat1.add_nuclide(openmc.Nuclide('U-235', '71c'), 1.0)
|
||||
materials = openmc.MaterialsFile()
|
||||
materials = openmc.Materials()
|
||||
materials.add_material(mat1)
|
||||
materials.export_to_xml()
|
||||
|
||||
|
|
@ -31,9 +29,7 @@ class SourceTestHarness(PyAPITestHarness):
|
|||
root.add_cell(inside_sphere)
|
||||
geometry = openmc.Geometry()
|
||||
geometry.root_universe = root
|
||||
geometry_xml = openmc.GeometryFile()
|
||||
geometry_xml.geometry = geometry
|
||||
geometry_xml.export_to_xml()
|
||||
geometry.export_to_xml()
|
||||
|
||||
# Create an array of different sources
|
||||
x_dist = openmc.stats.Uniform(-3., 3.)
|
||||
|
|
@ -56,11 +52,11 @@ class SourceTestHarness(PyAPITestHarness):
|
|||
energy2 = openmc.stats.Watt(0.988, 2.249)
|
||||
energy3 = openmc.stats.Tabular(E, p, interpolation='histogram')
|
||||
|
||||
source1 = Source(spatial1, angle1, energy1, strength=0.5)
|
||||
source2 = Source(spatial2, angle2, energy2, strength=0.3)
|
||||
source3 = Source(spatial3, angle3, energy3, strength=0.2)
|
||||
source1 = openmc.Source(spatial1, angle1, energy1, strength=0.5)
|
||||
source2 = openmc.Source(spatial2, angle2, energy2, strength=0.3)
|
||||
source3 = openmc.Source(spatial3, angle3, energy3, strength=0.2)
|
||||
|
||||
settings = openmc.SettingsFile()
|
||||
settings = openmc.Settings()
|
||||
settings.batches = 10
|
||||
settings.inactive = 5
|
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settings.particles = 1000
|
||||
|
|
|
|||
|
|
@ -4,7 +4,7 @@ import os
|
|||
import sys
|
||||
sys.path.insert(0, os.pardir)
|
||||
from testing_harness import PyAPITestHarness
|
||||
from openmc import Filter, Mesh, Tally, TalliesFile
|
||||
from openmc import Filter, Mesh, Tally, Tallies
|
||||
from openmc.source import Source
|
||||
from openmc.stats import Box
|
||||
|
||||
|
|
@ -170,7 +170,7 @@ class TalliesTestHarness(PyAPITestHarness):
|
|||
all_nuclide_tallies[0].estimator = 'tracklength'
|
||||
all_nuclide_tallies[0].estimator = 'collision'
|
||||
|
||||
self._input_set.tallies = TalliesFile()
|
||||
self._input_set.tallies = Tallies()
|
||||
self._input_set.tallies.add_tally(azimuthal_tally1)
|
||||
self._input_set.tallies.add_tally(azimuthal_tally2)
|
||||
self._input_set.tallies.add_tally(azimuthal_tally3)
|
||||
|
|
|
|||
|
|
@ -16,7 +16,7 @@ class TallyAggregationTestHarness(PyAPITestHarness):
|
|||
self._input_set.settings.output = {'summary': True}
|
||||
|
||||
# Initialize the tallies file
|
||||
tallies_file = openmc.TalliesFile()
|
||||
tallies_file = openmc.Tallies()
|
||||
|
||||
# Initialize the nuclides
|
||||
u235 = openmc.Nuclide('U-235')
|
||||
|
|
|
|||
|
|
@ -16,7 +16,7 @@ class TallyArithmeticTestHarness(PyAPITestHarness):
|
|||
self._input_set.settings.output = {'summary': True}
|
||||
|
||||
# Initialize the tallies file
|
||||
tallies_file = openmc.TalliesFile()
|
||||
tallies_file = openmc.Tallies()
|
||||
|
||||
# Initialize the nuclides
|
||||
u235 = openmc.Nuclide('U-235')
|
||||
|
|
|
|||
|
|
@ -17,7 +17,7 @@ class TallySliceMergeTestHarness(PyAPITestHarness):
|
|||
self._input_set.settings.output = {'summary': True}
|
||||
|
||||
# Initialize the tallies file
|
||||
tallies_file = openmc.TalliesFile()
|
||||
tallies_file = openmc.Tallies()
|
||||
|
||||
# Define nuclides and scores to add to both tallies
|
||||
self.nuclides = ['U-235', 'U-238']
|
||||
|
|
@ -69,8 +69,8 @@ class TallySliceMergeTestHarness(PyAPITestHarness):
|
|||
for nuclide in self.nuclides:
|
||||
distribcell_tally.add_nuclide(nuclide)
|
||||
|
||||
# Add tallies to a TalliesFile
|
||||
tallies_file = openmc.TalliesFile()
|
||||
# Add tallies to a Tallies object
|
||||
tallies_file = openmc.Tallies()
|
||||
tallies_file.add_tally(tallies[0])
|
||||
tallies_file.add_tally(distribcell_tally)
|
||||
|
||||
|
|
@ -95,7 +95,7 @@ class TallySliceMergeTestHarness(PyAPITestHarness):
|
|||
|
||||
# Slice the tallies by cell filter bins
|
||||
cell_filter_prod = itertools.product(tallies, self.cell_filters)
|
||||
tallies = map(lambda tf: tf[0].get_slice(filters=[tf[1].type],
|
||||
tallies = map(lambda tf: tf[0].get_slice(filters=[tf[1].type],
|
||||
filter_bins=[tf[1].get_bin(0)]), cell_filter_prod)
|
||||
|
||||
# Slice the tallies by energy filter bins
|
||||
|
|
@ -133,11 +133,11 @@ class TallySliceMergeTestHarness(PyAPITestHarness):
|
|||
# Extract the distribcell tally
|
||||
distribcell_tally = sp.get_tally(name='distribcell tally')
|
||||
|
||||
# Sum up a few subdomains from the distribcell tally
|
||||
sum1 = distribcell_tally.summation(filter_type='distribcell',
|
||||
# Sum up a few subdomains from the distribcell tally
|
||||
sum1 = distribcell_tally.summation(filter_type='distribcell',
|
||||
filter_bins=[0,100,2000,30000])
|
||||
# Sum up a few subdomains from the distribcell tally
|
||||
sum2 = distribcell_tally.summation(filter_type='distribcell',
|
||||
sum2 = distribcell_tally.summation(filter_type='distribcell',
|
||||
filter_bins=[500,5000,50000])
|
||||
|
||||
# Merge the distribcell tally slices
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue