Update use of Python API in C5G7 scripts

mit-crpg · May 26, 2016 · 468e49c · 468e49c
1 parent 0ed38d8
commit 468e49c
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Showing 10 changed files with 506 additions and 165 deletions.
diff --git a/.gitignore b/.gitignore
@@ -16,3 +16,6 @@ comou?
 
 # Compiled python files
 *.pyc
+
+# Python API-generated XML files
+c5g7/openmc/**/*.xml
diff --git a/c5g7/openmc/2d/build-xml-2d.py b/c5g7/openmc/2d/build-xml-2d.py
@@ -1,6 +1,4 @@
 import openmc
-from openmc.source import Source
-from openmc.stats import Box
 import sys
 sys.path.append('../')
 from lattices import lattices, universes, cells, surfaces
@@ -22,10 +20,9 @@
 
 from materials import materials
 
-# Instantiate a MaterialsFile, register all Materials, and export to XML
-materials_file = openmc.MaterialsFile()
+# Instantiate a Materials collection, register all Materials, and export to XML
+materials_file = openmc.Materials(materials.values())
 materials_file.default_xs = '300k'
-materials_file.add_materials(materials.values())
 materials_file.export_to_xml()
 
 
@@ -47,30 +44,26 @@
 lattices['Core'].universes = [[u, m, w], [m, u, w], [w, w, w]]
 cells['Core'].fill = lattices['Core']
 
-# Instantiate a Geometry and register the root Universe
+# Instantiate a Geometry, register the root Universe, and export to XML
 geometry = openmc.Geometry()
 geometry.root_universe = universes['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
 ###############################################################################
 
-# Instantiate a SettingsFile, set all runtime parameters, and export to XML
-settings_file = openmc.SettingsFile()
+# Instantiate a Settings, 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
 settings_file.inactive = inactive
 settings_file.particles = particles
 settings_file.output = {'tallies': True, 'summary': True}
-settings_file.source = Source(space=Box([-32.13, -10.71, -1.0],
-                                        [10.71, 32.13, 1.0]))
+settings_file.source = openmc.Source(space=openmc.stats.Box(
+    [-32.13, -10.71, -1.0], [10.71, 32.13, 1.0]))
 settings_file.entropy_lower_left = [-32.13, -32.13, -1.E50]
 settings_file.entropy_upper_right = [32.13, 32.13, 1.E50]
 settings_file.entropy_dimension = [51, 51, 1]
@@ -89,20 +82,14 @@
 plot_1.color = 'mat'
 plot_1.basis = 'xy'
 
-# Instantiate a PlotsFile, add Plot, and export to XML
-plot_file = openmc.PlotsFile()
-plot_file.add_plot(plot_1)
+# Instantiate a Plots collection and export to XML
+plot_file = openmc.Plots([plot_1])
 plot_file.export_to_xml()
 
 ###############################################################################
 #                   Exporting to OpenMC tallies.xml File
 ###############################################################################
 
-# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
-tallies_file = openmc.TalliesFile()
-tallies_file.add_mesh(mesh)
-
-for tally in tallies.values():
-    tallies_file.add_tally(tally)
-
+# Instantiate a Tallies, register Tally/Mesh, and export to XML
+tallies_file = openmc.Tallies(tallies.values())
 tallies_file.export_to_xml()
diff --git a/c5g7/openmc/3d/build-xml-3d.py b/c5g7/openmc/3d/build-xml-3d.py
@@ -1,6 +1,4 @@
 import openmc
-from openmc.source import Source
-from openmc.stats import Box
 import sys
 sys.path.append('../')
 from lattices import lattices, universes, cells, surfaces
@@ -22,10 +20,9 @@
 
 from materials import materials
 
-# Instantiate a MaterialsFile, register all Materials, and export to XML
-materials_file = openmc.MaterialsFile()
+# Instantiate a Materials collection, register all Materials, and export to XML
+materials_file = openmc.Materials(materials.values())
 materials_file.default_xs = '300k'
-materials_file.add_materials(materials.values())
 materials_file.export_to_xml()
 
 
@@ -51,29 +48,26 @@
 # Add Core lattice to Core cell
 cells['Core'].fill = lattices['Core']
 
-# Instantiate a Geometry and register the root Universe
+# Instantiate a Geometry, register the root Universe, and export to XML
 geometry = openmc.Geometry()
 geometry.root_universe = universes['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
 ###############################################################################
 
-# Instantiate a SettingsFile, set all runtime parameters, and export to XML
-settings_file = openmc.SettingsFile()
+# Instantiate a Settings, 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
 settings_file.inactive = inactive
 settings_file.particles = particles
 settings_file.output = {'tallies': True, 'summary': True}
-settings_file.source = Source(space=Box([-32.13, -10.71, -107.1], [10.71, 32.13, 85.68]))
+settings_file.source = openmc.Source(space=openmc.stats.Box(
+    [-32.13, -10.71, -107.1], [10.71, 32.13, 85.68]))
 settings_file.entropy_lower_left = [-32.13, -32.13, -107.1]
 settings_file.entropy_upper_right = [32.13,  32.13,  107.1]
 settings_file.entropy_dimension = [51, 51, 30]
@@ -100,21 +94,14 @@
 plot_2.color = 'mat'
 plot_2.basis = 'xz'
 
-# Instantiate a PlotsFile, add Plot, and export to XML
-plot_file = openmc.PlotsFile()
-plot_file.add_plot(plot_1)
-plot_file.add_plot(plot_2)
+# Instantiate a Plots collection and export to XML
+plot_file = openmc.Plots([plot_1, plot_2])
 plot_file.export_to_xml()
 
 ###############################################################################
 #                   Exporting to OpenMC tallies.xml File
 ###############################################################################
 
-# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
-tallies_file = openmc.TalliesFile()
-tallies_file.add_mesh(mesh)
-
-for tally in tallies.values():
-    tallies_file.add_tally(tally)
-
+# Instantiate a Tallies collection and export to XML
+tallies_file = openmc.Tallies(tallies.values())
 tallies_file.export_to_xml()
diff --git a/c5g7/openmc/materials.py b/c5g7/openmc/materials.py
@@ -130,7 +130,7 @@
 control_rod_xsdata.nu_fission = np.zeros(7)
 control_rod_xsdata.chi = np.zeros(7)
 
-mg_cross_sections_file = openmc.MGXSLibraryFile(groups)
+mg_cross_sections_file = openmc.MGXSLibrary(groups)
 mg_cross_sections_file.add_xsdatas([uo2_xsdata, mox43_xsdata, mox7_xsdata, mox87_xsdata,
                                     fiss_chamber_xsdata, guide_tube_xsdata, water_xsdata,
                                     control_rod_xsdata])

diff --git a/c5g7/openmc/pin/build-xml-pin.py b/c5g7/openmc/pin/build-xml-pin.py
@@ -1,6 +1,4 @@
 import openmc
-from openmc.source import Source
-from openmc.stats import Box
 import sys
 sys.path.append('../')
 import cells
@@ -21,10 +19,9 @@
 
 from materials import materials
 
-# Instantiate a MaterialsFile, register all Materials, and export to XML
-materials_file = openmc.MaterialsFile()
+# Instantiate a Materials, register all Materials, and export to XML
+materials_file = openmc.Materials(materials.values())
 materials_file.default_xs = '300k'
-materials_file.add_materials(materials.values())
 materials_file.export_to_xml()
 
 
@@ -37,31 +34,26 @@
 universes['Root'] = openmc.Universe(universe_id=0,  name='Root')
 universes['Root'].add_cells([cells.cells['UO2'], cells.cells['UO2 Pin']])
 
-# Instantiate a Geometry and register the root Universe
+# Instantiate a Geometry, register the root Universe, and export to XML
 geometry = openmc.Geometry()
 geometry.root_universe = universes['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
 ###############################################################################
 
-# Instantiate a SettingsFile, set all runtime parameters, and export to XML
-settings_file = openmc.SettingsFile()
+# Instantiate a Settings, 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
 settings_file.inactive = inactive
 settings_file.particles = particles
 settings_file.output = {'tallies': True, 'summary': True}
-settings_file.source = Source(space=Box([-0.63,-0.63,-1.E50],
-                                        [ 0.63, 0.63, 1.E50],
-                                        only_fissionable=True))
+settings_file.source = openmc.Source(space=openmc.stats.Box(
+    [-0.63, -0.63, -1.e50], [0.63, 0.63, 1.e50], only_fissionable=True))
 settings_file.entropy_lower_left  = [-0.63,-0.63,-1.E50]
 settings_file.entropy_upper_right = [ 0.63, 0.63, 1.E50]
 settings_file.entropy_dimension = [10,10,1]
@@ -80,9 +72,8 @@
 plot_1.color = 'mat'
 plot_1.basis = 'xy'
 
-# Instantiate a PlotsFile, add Plot, and export to XML
-plot_file = openmc.PlotsFile()
-plot_file.add_plot(plot_1)
+# Instantiate a Plots collection and export to XML
+plot_file = openmc.PlotsFile([plot_1])
 plot_file.export_to_xml()
 
 ###############################################################################
@@ -105,21 +96,12 @@
 
 # Instantiate the Tally
 tallies['Mesh Rates'] = openmc.Tally(tally_id=1, name='tally 1')
-tallies['Mesh Rates'].add_filter(mesh_filter)
-tallies['Mesh Rates'].add_score('flux')
-tallies['Mesh Rates'].add_score('fission')
-tallies['Mesh Rates'].add_score('nu-fission')
+tallies['Mesh Rates'].filters = [mesh_filter]
+tallies['Mesh Rates'].scores = ['flux', 'fission', 'nu-fission']
 
 tallies['Global Rates'] = openmc.Tally(tally_id=2, name='tally 2')
-tallies['Global Rates'].add_score('flux')
-tallies['Global Rates'].add_score('fission')
-tallies['Global Rates'].add_score('nu-fission')
-
-# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
-tallies_file = openmc.TalliesFile()
-tallies_file.add_mesh(mesh)
-
-for tally in tallies.values():
-    tallies_file.add_tally(tally)
+tallies['Global Rates'].scores = ['flux', 'fission', 'nu-fission']
 
+# Instantiate a Tallies collection and export to XML
+tallies_file = openmc.Tallies(tallies.values())
 tallies_file.export_to_xml()
diff --git a/c5g7/openmc/rodded-A/build-xml-rodded-A.py b/c5g7/openmc/rodded-A/build-xml-rodded-A.py
@@ -1,6 +1,4 @@
 import openmc
-from openmc.source import Source
-from openmc.stats import Box
 import sys
 sys.path.append('../')
 from lattices import lattices, universes, cells, surfaces
@@ -22,10 +20,9 @@
 
 from materials import materials
 
-# Instantiate a MaterialsFile, register all Materials, and export to XML
-materials_file = openmc.MaterialsFile()
+# Instantiate a Materials collection and export to XML
+materials_file = openmc.Materials(materials.values())
 materials_file.default_xs = '300k'
-materials_file.add_materials(materials.values())
 materials_file.export_to_xml()
 
 
@@ -55,29 +52,26 @@
 # Add Core lattice to Core cell
 cells['Core'].fill = lattices['Core']
 
-# Instantiate a Geometry and register the root Universe
+# Instantiate a Geometry, register the root Universe, and export to XML
 geometry = openmc.Geometry()
 geometry.root_universe = universes['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
 ###############################################################################
 
-# Instantiate a SettingsFile, set all runtime parameters, and export to XML
-settings_file = openmc.SettingsFile()
+# Instantiate a Settings, 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
 settings_file.inactive = inactive
 settings_file.particles = particles
 settings_file.output = {'tallies': True, 'summary': True}
-settings_file.source = Source(space=Box([-32.13, -10.71, -107.1], [10.71, 32.13, 85.68]))
+settings_file.source = openmc.Source(space=openmc.stats.Box(
+    [-32.13, -10.71, -107.1], [10.71, 32.13, 85.68]))
 settings_file.entropy_lower_left = [-32.13, -32.13, -107.1]
 settings_file.entropy_upper_right = [32.13,  32.13,  107.1]
 settings_file.entropy_dimension = [51, 51, 30]
@@ -104,21 +98,14 @@
 plot_2.color = 'mat'
 plot_2.basis = 'xz'
 
-# Instantiate a PlotsFile, add Plot, and export to XML
-plot_file = openmc.PlotsFile()
-plot_file.add_plot(plot_1)
-plot_file.add_plot(plot_2)
+# Instantiate a Plots collection and export to XML
+plot_file = openmc.Plots([plot_1, plot_2])
 plot_file.export_to_xml()
 
 ###############################################################################
 #                   Exporting to OpenMC tallies.xml File
 ###############################################################################
 
-# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
-tallies_file = openmc.TalliesFile()
-tallies_file.add_mesh(mesh)
-
-for tally in tallies.values():
-    tallies_file.add_tally(tally)
-
+# Instantiate a Tallies and export to XML
+tallies_file = openmc.Tallies(tallies.values())
 tallies_file.export_to_xml()