diff --git a/examples/python/basic/build-xml.py b/examples/python/basic/build-xml.py index 18a16aab6..58825ee49 100644 --- a/examples/python/basic/build-xml.py +++ b/examples/python/basic/build-xml.py @@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235') # Instantiate some Materials and register the appropriate Nuclides moderator = openmc.Material(material_id=41, name='moderator') -moderator.setDensity('g/cc', 1.0) -moderator.addNuclide(h1, 2.) -moderator.addNuclide(o16, 1.) -moderator.addSAlphaBeta('HH2O', '71t') +moderator.set_density('g/cc', 1.0) +moderator.add_nuclide(h1, 2.) +moderator.add_nuclide(o16, 1.) +moderator.add_s_alpha_beta('HH2O', '71t') fuel = openmc.Material(material_id=40, name='fuel') -fuel.setDensity('g/cc', 4.5) -fuel.addNuclide(u235, 1.) +fuel.set_density('g/cc', 4.5) +fuel.add_nuclide(u235, 1.) # Instantiate a MaterialsFile, register all Materials, and export to XML materials_file = openmc.MaterialsFile() -materials_file.setDefaultXS('71c') -materials_file.addMaterials([moderator, fuel]) -materials_file.exportToXML() +materials_file.set_default_xs('71c') +materials_file.add_materials([moderator, fuel]) +materials_file.export_to_xml() ############################################################################### @@ -46,7 +46,7 @@ materials_file.exportToXML() surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=7, name='surf 1') surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=9, name='surf 2') surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=11, name='surf 3') -surf3.setBoundaryType('vacuum') +surf3.set_boundary_type('vacuum') # Instantiate Cells cell1 = openmc.Cell(cell_id=1, name='cell 1') @@ -55,34 +55,34 @@ cell3 = openmc.Cell(cell_id=101, name='cell 3') cell4 = openmc.Cell(cell_id=2, name='cell 4') # Register Surfaces with Cells -cell1.addSurface(surface=surf2, halfspace=-1) -cell2.addSurface(surface=surf1, halfspace=-1) -cell3.addSurface(surface=surf1, halfspace=+1) -cell4.addSurface(surface=surf2, halfspace=+1) -cell4.addSurface(surface=surf3, halfspace=-1) +cell1.add_surface(surface=surf2, halfspace=-1) +cell2.add_surface(surface=surf1, halfspace=-1) +cell3.add_surface(surface=surf1, halfspace=+1) +cell4.add_surface(surface=surf2, halfspace=+1) +cell4.add_surface(surface=surf3, halfspace=-1) # Register Materials with Cells -cell2.setFill(fuel) -cell3.setFill(moderator) -cell4.setFill(moderator) +cell2.set_fill(fuel) +cell3.set_fill(moderator) +cell4.set_fill(moderator) # Instantiate Universes universe1 = openmc.Universe(universe_id=37) root = openmc.Universe(universe_id=0, name='root universe') -cell1.setFill(universe1) +cell1.set_fill(universe1) # Register Cells with Universes -universe1.addCells([cell2, cell3]) -root.addCells([cell1, cell4]) +universe1.add_cells([cell2, cell3]) +root.add_cells([cell1, cell4]) # Instantiate a Geometry and register the root Universe geometry = openmc.Geometry() -geometry.setRootUniverse(root) +geometry.set_root_universe(root) # Instantiate a GeometryFile, register Geometry, and export to XML geometry_file = openmc.GeometryFile() -geometry_file.setGeometry(geometry) -geometry_file.exportToXML() +geometry_file.set_geometry(geometry) +geometry_file.export_to_xml() ############################################################################### @@ -91,11 +91,11 @@ geometry_file.exportToXML() # Instantiate a SettingsFile, set all runtime parameters, and export to XML settings_file = openmc.SettingsFile() -settings_file.setBatches(batches) -settings_file.setInactive(inactive) -settings_file.setParticles(particles) -settings_file.setSourceSpace('box', [-4, -4, -4, 4, 4, 4]) -settings_file.exportToXML() +settings_file.set_batches(batches) +settings_file.set_inactive(inactive) +settings_file.set_particles(particles) +settings_file.set_source_space('box', [-4, -4, -4, 4, 4, 4]) +settings_file.export_to_xml() ############################################################################### @@ -109,33 +109,33 @@ energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.]) # Instantiate the first Tally first_tally = openmc.Tally(tally_id=1, label='first tally') -first_tally.addFilter(cell_filter) +first_tally.add_filter(cell_filter) scores = ['total', 'scatter', 'nu-scatter', \ 'absorption', 'fission', 'nu-fission'] for score in scores: - first_tally.addScore(score) + first_tally.add_score(score) # Instantiate the second Tally second_tally = openmc.Tally(tally_id=2, label='second tally') -second_tally.addFilter(cell_filter) -second_tally.addFilter(energy_filter) +second_tally.add_filter(cell_filter) +second_tally.add_filter(energy_filter) scores = ['total', 'scatter', 'nu-scatter', \ 'absorption', 'fission', 'nu-fission'] for score in scores: - second_tally.addScore(score) + second_tally.add_score(score) # Instantiate the third Tally third_tally = openmc.Tally(tally_id=3, label='third tally') -third_tally.addFilter(cell_filter) -third_tally.addFilter(energy_filter) -third_tally.addFilter(energyout_filter) +third_tally.add_filter(cell_filter) +third_tally.add_filter(energy_filter) +third_tally.add_filter(energyout_filter) scores = ['scatter', 'nu-scatter', 'nu-fission'] for score in scores: - third_tally.addScore(score) + third_tally.add_score(score) # Instantiate a TalliesFile, register all Tallies, and export to XML tallies_file = openmc.TalliesFile() -tallies_file.addTally(first_tally) -tallies_file.addTally(second_tally) -tallies_file.addTally(third_tally) -tallies_file.exportToXML() +tallies_file.add_tally(first_tally) +tallies_file.add_tally(second_tally) +tallies_file.add_tally(third_tally) +tallies_file.export_to_xml() diff --git a/examples/python/lattice/nested/build-xml.py b/examples/python/lattice/nested/build-xml.py index b7d314f3f..2ad94a65c 100644 --- a/examples/python/lattice/nested/build-xml.py +++ b/examples/python/lattice/nested/build-xml.py @@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235') # Instantiate some Materials and register the appropriate Nuclides fuel = openmc.Material(material_id=1, name='fuel') -fuel.setDensity('g/cc', 4.5) -fuel.addNuclide(u235, 1.) +fuel.set_density('g/cc', 4.5) +fuel.add_nuclide(u235, 1.) moderator = openmc.Material(material_id=2, name='moderator') -moderator.setDensity('g/cc', 1.0) -moderator.addNuclide(h1, 2.) -moderator.addNuclide(o16, 1.) -moderator.addSAlphaBeta('HH2O', '71t') +moderator.set_density('g/cc', 1.0) +moderator.add_nuclide(h1, 2.) +moderator.add_nuclide(o16, 1.) +moderator.add_s_alpha_beta('HH2O', '71t') # Instantiate a MaterialsFile, register all Materials, and export to XML materials_file = openmc.MaterialsFile() -materials_file.setDefaultXS('71c') -materials_file.addMaterials([moderator, fuel]) -materials_file.exportToXML() +materials_file.set_default_xs('71c') +materials_file.add_materials([moderator, fuel]) +materials_file.export_to_xml() ############################################################################### @@ -51,10 +51,10 @@ fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4) fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3) fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2) -left.setBoundaryType('vacuum') -right.setBoundaryType('vacuum') -top.setBoundaryType('vacuum') -bottom.setBoundaryType('vacuum') +left.set_boundary_type('vacuum') +right.set_boundary_type('vacuum') +top.set_boundary_type('vacuum') +bottom.set_boundary_type('vacuum') # Instantiate Cells cell1 = openmc.Cell(cell_id=1, name='Cell 1') @@ -67,28 +67,28 @@ cell7 = openmc.Cell(cell_id=301, name='cell 7') cell8 = openmc.Cell(cell_id=302, name='cell 8') # Register Surfaces with Cells -cell1.addSurface(left, halfspace=+1) -cell1.addSurface(right, halfspace=-1) -cell1.addSurface(bottom, halfspace=+1) -cell1.addSurface(top, halfspace=-1) -cell2.addSurface(left, halfspace=+1) -cell2.addSurface(right, halfspace=-1) -cell2.addSurface(bottom, halfspace=+1) -cell2.addSurface(top, halfspace=-1) -cell3.addSurface(fuel1, halfspace=-1) -cell4.addSurface(fuel1, halfspace=+1) -cell5.addSurface(fuel2, halfspace=-1) -cell6.addSurface(fuel2, halfspace=+1) -cell7.addSurface(fuel3, halfspace=-1) -cell8.addSurface(fuel3, halfspace=+1) +cell1.add_surface(left, halfspace=+1) +cell1.add_surface(right, halfspace=-1) +cell1.add_surface(bottom, halfspace=+1) +cell1.add_surface(top, halfspace=-1) +cell2.add_surface(left, halfspace=+1) +cell2.add_surface(right, halfspace=-1) +cell2.add_surface(bottom, halfspace=+1) +cell2.add_surface(top, halfspace=-1) +cell3.add_surface(fuel1, halfspace=-1) +cell4.add_surface(fuel1, halfspace=+1) +cell5.add_surface(fuel2, halfspace=-1) +cell6.add_surface(fuel2, halfspace=+1) +cell7.add_surface(fuel3, halfspace=-1) +cell8.add_surface(fuel3, halfspace=+1) # Register Materials with Cells -cell3.setFill(fuel) -cell4.setFill(moderator) -cell5.setFill(fuel) -cell6.setFill(moderator) -cell7.setFill(fuel) -cell8.setFill(moderator) +cell3.set_fill(fuel) +cell4.set_fill(moderator) +cell5.set_fill(fuel) +cell6.set_fill(moderator) +cell7.set_fill(fuel) +cell8.set_fill(moderator) # Instantiate Universe univ1 = openmc.Universe(universe_id=1) @@ -98,39 +98,39 @@ univ4 = openmc.Universe(universe_id=5) root = openmc.Universe(universe_id=0, name='root universe') # Register Cells with Universe -univ1.addCells([cell3, cell4]) -univ2.addCells([cell5, cell6]) -univ3.addCells([cell7, cell8]) -root.addCell(cell1) -univ4.addCell(cell2) +univ1.add_cells([cell3, cell4]) +univ2.add_cells([cell5, cell6]) +univ3.add_cells([cell7, cell8]) +root.add_cell(cell1) +univ4.add_cell(cell2) # Instantiate nested Lattices lattice1 = openmc.Lattice(lattice_id=4, name='4x4 assembly') -lattice1.setDimension([2, 2]) -lattice1.setLowerLeft([-1., -1.]) -lattice1.setWidth([1., 1.]) -lattice1.setUniverses([[univ1, univ2], +lattice1.set_dimension([2, 2]) +lattice1.set_lower_left([-1., -1.]) +lattice1.set_width([1., 1.]) +lattice1.set_universes([[univ1, univ2], [univ2, univ3]]) lattice2 = openmc.Lattice(lattice_id=6, name='4x4 core') -lattice2.setDimension([2, 2]) -lattice2.setLowerLeft([-2., -2.]) -lattice2.setWidth([2., 2.]) -lattice2.setUniverses([[univ4, univ4], +lattice2.set_dimension([2, 2]) +lattice2.set_lower_left([-2., -2.]) +lattice2.set_width([2., 2.]) +lattice2.set_universes([[univ4, univ4], [univ4, univ4]]) # Fill Cell with the Lattice -cell1.setFill(lattice2) -cell2.setFill(lattice1) +cell1.set_fill(lattice2) +cell2.set_fill(lattice1) # Instantiate a Geometry and register the root Universe geometry = openmc.Geometry() -geometry.setRootUniverse(root) +geometry.set_root_universe(root) # Instantiate a GeometryFile, register Geometry, and export to XML geometry_file = openmc.GeometryFile() -geometry_file.setGeometry(geometry) -geometry_file.exportToXML() +geometry_file.set_geometry(geometry) +geometry_file.export_to_xml() ############################################################################### @@ -139,11 +139,11 @@ geometry_file.exportToXML() # Instantiate a SettingsFile, set all runtime parameters, and export to XML settings_file = openmc.SettingsFile() -settings_file.setBatches(batches) -settings_file.setInactive(inactive) -settings_file.setParticles(particles) -settings_file.setSourceSpace('box', [-1, -1, -1, 1, 1, 1]) -settings_file.exportToXML() +settings_file.set_batches(batches) +settings_file.set_inactive(inactive) +settings_file.set_particles(particles) +settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1]) +settings_file.export_to_xml() ############################################################################### @@ -151,15 +151,15 @@ settings_file.exportToXML() ############################################################################### plot = openmc.Plot(plot_id=1) -plot.setOrigin([0, 0, 0]) -plot.setWidth([4, 4]) -plot.setPixels([400, 400]) -plot.setColor('mat') +plot.set_origin([0, 0, 0]) +plot.set_width([4, 4]) +plot.set_pixels([400, 400]) +plot.set_color('mat') # Instantiate a PlotsFile, add Plot, and export to XML plot_file = openmc.PlotsFile() -plot_file.addPlot(plot) -plot_file.exportToXML() +plot_file.add_plot(plot) +plot_file.export_to_xml() ############################################################################### @@ -168,22 +168,22 @@ plot_file.exportToXML() # Instantiate a tally mesh mesh = openmc.Mesh(mesh_id=1) -mesh.setType('rectangular') -mesh.setDimension([4, 4]) -mesh.setLowerLeft([-2, -2]) -mesh.setWidth([1, 1]) +mesh.set_type('rectangular') +mesh.set_dimension([4, 4]) +mesh.set_lower_left([-2, -2]) +mesh.set_width([1, 1]) # Instantiate tally Filter mesh_filter = openmc.Filter() -mesh_filter.setMesh(mesh) +mesh_filter.set_mesh(mesh) # Instantiate the Tally tally = openmc.Tally(tally_id=1) -tally.addFilter(mesh_filter) -tally.addScore('total') +tally.add_filter(mesh_filter) +tally.add_score('total') # Instantiate a TalliesFile, register Tally/Mesh, and export to XML tallies_file = openmc.TalliesFile() -tallies_file.addMesh(mesh) -tallies_file.addTally(tally) -tallies_file.exportToXML() +tallies_file.add_mesh(mesh) +tallies_file.add_tally(tally) +tallies_file.export_to_xml() diff --git a/examples/python/lattice/simple/build-xml.py b/examples/python/lattice/simple/build-xml.py index 9c0dbf03d..590441aee 100644 --- a/examples/python/lattice/simple/build-xml.py +++ b/examples/python/lattice/simple/build-xml.py @@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235') # Instantiate some Materials and register the appropriate Nuclides fuel = openmc.Material(material_id=1, name='fuel') -fuel.setDensity('g/cc', 4.5) -fuel.addNuclide(u235, 1.) +fuel.set_density('g/cc', 4.5) +fuel.add_nuclide(u235, 1.) moderator = openmc.Material(material_id=2, name='moderator') -moderator.setDensity('g/cc', 1.0) -moderator.addNuclide(h1, 2.) -moderator.addNuclide(o16, 1.) -moderator.addSAlphaBeta('HH2O', '71t') +moderator.set_density('g/cc', 1.0) +moderator.add_nuclide(h1, 2.) +moderator.add_nuclide(o16, 1.) +moderator.add_s_alpha_beta('HH2O', '71t') # Instantiate a MaterialsFile, register all Materials, and export to XML materials_file = openmc.MaterialsFile() -materials_file.setDefaultXS('71c') -materials_file.addMaterials([moderator, fuel]) -materials_file.exportToXML() +materials_file.set_default_xs('71c') +materials_file.add_materials([moderator, fuel]) +materials_file.export_to_xml() ############################################################################### @@ -51,10 +51,10 @@ fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4) fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3) fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2) -left.setBoundaryType('vacuum') -right.setBoundaryType('vacuum') -top.setBoundaryType('vacuum') -bottom.setBoundaryType('vacuum') +left.set_boundary_type('vacuum') +right.set_boundary_type('vacuum') +top.set_boundary_type('vacuum') +bottom.set_boundary_type('vacuum') # Instantiate Cells cell1 = openmc.Cell(cell_id=1, name='Cell 1') @@ -66,24 +66,24 @@ cell6 = openmc.Cell(cell_id=301, name='cell 6') cell7 = openmc.Cell(cell_id=302, name='cell 7') # Register Surfaces with Cells -cell1.addSurface(left, halfspace=+1) -cell1.addSurface(right, halfspace=-1) -cell1.addSurface(bottom, halfspace=+1) -cell1.addSurface(top, halfspace=-1) -cell2.addSurface(fuel1, halfspace=-1) -cell3.addSurface(fuel1, halfspace=+1) -cell4.addSurface(fuel2, halfspace=-1) -cell5.addSurface(fuel2, halfspace=+1) -cell6.addSurface(fuel3, halfspace=-1) -cell7.addSurface(fuel3, halfspace=+1) +cell1.add_surface(left, halfspace=+1) +cell1.add_surface(right, halfspace=-1) +cell1.add_surface(bottom, halfspace=+1) +cell1.add_surface(top, halfspace=-1) +cell2.add_surface(fuel1, halfspace=-1) +cell3.add_surface(fuel1, halfspace=+1) +cell4.add_surface(fuel2, halfspace=-1) +cell5.add_surface(fuel2, halfspace=+1) +cell6.add_surface(fuel3, halfspace=-1) +cell7.add_surface(fuel3, halfspace=+1) # Register Materials with Cells -cell2.setFill(fuel) -cell3.setFill(moderator) -cell4.setFill(fuel) -cell5.setFill(moderator) -cell6.setFill(fuel) -cell7.setFill(moderator) +cell2.set_fill(fuel) +cell3.set_fill(moderator) +cell4.set_fill(fuel) +cell5.set_fill(moderator) +cell6.set_fill(fuel) +cell7.set_fill(moderator) # Instantiate Universe univ1 = openmc.Universe(universe_id=1) @@ -92,32 +92,32 @@ univ3 = openmc.Universe(universe_id=3) root = openmc.Universe(universe_id=0, name='root universe') # Register Cells with Universe -univ1.addCells([cell2, cell3]) -univ2.addCells([cell4, cell5]) -univ3.addCells([cell6, cell7]) -root.addCell(cell1) +univ1.add_cells([cell2, cell3]) +univ2.add_cells([cell4, cell5]) +univ3.add_cells([cell6, cell7]) +root.add_cell(cell1) # Instantiate a Lattice lattice = openmc.Lattice(lattice_id=5) -lattice.setDimension([4, 4]) -lattice.setLowerLeft([-2., -2.]) -lattice.setWidth([1., 1.]) -lattice.setUniverses([[univ1, univ2, univ1, univ2], +lattice.set_dimension([4, 4]) +lattice.set_lower_left([-2., -2.]) +lattice.set_width([1., 1.]) +lattice.set_universes([[univ1, univ2, univ1, univ2], [univ2, univ3, univ2, univ3], [univ1, univ2, univ1, univ2], [univ2, univ3, univ2, univ3]]) # Fill Cell with the Lattice -cell1.setFill(lattice) +cell1.set_fill(lattice) # Instantiate a Geometry and register the root Universe geometry = openmc.Geometry() -geometry.setRootUniverse(root) +geometry.set_root_universe(root) # Instantiate a GeometryFile, register Geometry, and export to XML geometry_file = openmc.GeometryFile() -geometry_file.setGeometry(geometry) -geometry_file.exportToXML() +geometry_file.set_geometry(geometry) +geometry_file.export_to_xml() ############################################################################### @@ -126,11 +126,11 @@ geometry_file.exportToXML() # Instantiate a SettingsFile, set all runtime parameters, and export to XML settings_file = openmc.SettingsFile() -settings_file.setBatches(batches) -settings_file.setInactive(inactive) -settings_file.setParticles(particles) -settings_file.setSourceSpace('box', [-1, -1, -1, 1, 1, 1]) -settings_file.exportToXML() +settings_file.set_batches(batches) +settings_file.set_inactive(inactive) +settings_file.set_particles(particles) +settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1]) +settings_file.export_to_xml() ############################################################################### @@ -138,15 +138,15 @@ settings_file.exportToXML() ############################################################################### plot = openmc.Plot(plot_id=1) -plot.setOrigin([0, 0, 0]) -plot.setWidth([4, 4]) -plot.setPixels([400, 400]) -plot.setColor('mat') +plot.set_origin([0, 0, 0]) +plot.set_width([4, 4]) +plot.set_pixels([400, 400]) +plot.set_color('mat') # Instantiate a PlotsFile, add Plot, and export to XML plot_file = openmc.PlotsFile() -plot_file.addPlot(plot) -plot_file.exportToXML() +plot_file.add_plot(plot) +plot_file.export_to_xml() ############################################################################### @@ -155,22 +155,22 @@ plot_file.exportToXML() # Instantiate a tally mesh mesh = openmc.Mesh(mesh_id=1) -mesh.setType('rectangular') -mesh.setDimension([4, 4]) -mesh.setLowerLeft([-2, -2]) -mesh.setWidth([1, 1]) +mesh.set_type('rectangular') +mesh.set_dimension([4, 4]) +mesh.set_lower_left([-2, -2]) +mesh.set_width([1, 1]) # Instantiate tally Filter mesh_filter = openmc.Filter() -mesh_filter.setMesh(mesh) +mesh_filter.set_mesh(mesh) # Instantiate the Tally tally = openmc.Tally(tally_id=1) -tally.addFilter(mesh_filter) -tally.addScore('total') +tally.add_filter(mesh_filter) +tally.add_score('total') # Instantiate a TalliesFile, register Tally/Mesh, and export to XML tallies_file = openmc.TalliesFile() -tallies_file.addMesh(mesh) -tallies_file.addTally(tally) -tallies_file.exportToXML() +tallies_file.add_mesh(mesh) +tallies_file.add_tally(tally) +tallies_file.export_to_xml() diff --git a/examples/python/pincell/build-xml.py b/examples/python/pincell/build-xml.py index 6916bef9c..4da66fe45 100644 --- a/examples/python/pincell/build-xml.py +++ b/examples/python/pincell/build-xml.py @@ -52,60 +52,60 @@ u238 = openmc.Nuclide('U-238') # Instantiate some Materials and register the appropriate Nuclides uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment') -uo2.setDensity('g/cm3', 10.29769) -uo2.addNuclide(u234, 4.4843e-6) -uo2.addNuclide(u235, 5.5815e-4) -uo2.addNuclide(u238, 2.2408e-2) -uo2.addNuclide(o16, 4.5829e-2) -uo2.addNuclide(o17, 1.1164e-4) +uo2.set_density('g/cm3', 10.29769) +uo2.add_nuclide(u234, 4.4843e-6) +uo2.add_nuclide(u235, 5.5815e-4) +uo2.add_nuclide(u238, 2.2408e-2) +uo2.add_nuclide(o16, 4.5829e-2) +uo2.add_nuclide(o17, 1.1164e-4) helium = openmc.Material(material_id=2, name='Helium for gap') -helium.setDensity('g/cm3', 0.001598) -helium.addNuclide(he4, 2.4044e-4) +helium.set_density('g/cm3', 0.001598) +helium.add_nuclide(he4, 2.4044e-4) zircaloy = openmc.Material(material_id=3, name='Zircaloy 4') -zircaloy.setDensity('g/cm3', 6.55) -zircaloy.addNuclide(o16, 3.0743e-4) -zircaloy.addNuclide(o17, 7.4887e-7) -zircaloy.addNuclide(cr50, 3.2962e-6) -zircaloy.addNuclide(cr52, 6.3564e-5) -zircaloy.addNuclide(cr53, 7.2076e-6) -zircaloy.addNuclide(cr54, 1.7941e-6) -zircaloy.addNuclide(fe54, 8.6699e-6) -zircaloy.addNuclide(fe56, 1.3610e-4) -zircaloy.addNuclide(fe57, 3.1431e-6) -zircaloy.addNuclide(fe58, 4.1829e-7) -zircaloy.addNuclide(zr90, 2.1827e-2) -zircaloy.addNuclide(zr91, 4.7600e-3) -zircaloy.addNuclide(zr92, 7.2758e-3) -zircaloy.addNuclide(zr94, 7.3734e-3) -zircaloy.addNuclide(zr96, 1.1879e-3) -zircaloy.addNuclide(sn112, 4.6735e-6) -zircaloy.addNuclide(sn114, 3.1799e-6) -zircaloy.addNuclide(sn115, 1.6381e-6) -zircaloy.addNuclide(sn116, 7.0055e-5) -zircaloy.addNuclide(sn117, 3.7003e-5) -zircaloy.addNuclide(sn118, 1.1669e-4) -zircaloy.addNuclide(sn119, 4.1387e-5) -zircaloy.addNuclide(sn120, 1.5697e-4) -zircaloy.addNuclide(sn122, 2.2308e-5) -zircaloy.addNuclide(sn124, 2.7897e-5) +zircaloy.set_density('g/cm3', 6.55) +zircaloy.add_nuclide(o16, 3.0743e-4) +zircaloy.add_nuclide(o17, 7.4887e-7) +zircaloy.add_nuclide(cr50, 3.2962e-6) +zircaloy.add_nuclide(cr52, 6.3564e-5) +zircaloy.add_nuclide(cr53, 7.2076e-6) +zircaloy.add_nuclide(cr54, 1.7941e-6) +zircaloy.add_nuclide(fe54, 8.6699e-6) +zircaloy.add_nuclide(fe56, 1.3610e-4) +zircaloy.add_nuclide(fe57, 3.1431e-6) +zircaloy.add_nuclide(fe58, 4.1829e-7) +zircaloy.add_nuclide(zr90, 2.1827e-2) +zircaloy.add_nuclide(zr91, 4.7600e-3) +zircaloy.add_nuclide(zr92, 7.2758e-3) +zircaloy.add_nuclide(zr94, 7.3734e-3) +zircaloy.add_nuclide(zr96, 1.1879e-3) +zircaloy.add_nuclide(sn112, 4.6735e-6) +zircaloy.add_nuclide(sn114, 3.1799e-6) +zircaloy.add_nuclide(sn115, 1.6381e-6) +zircaloy.add_nuclide(sn116, 7.0055e-5) +zircaloy.add_nuclide(sn117, 3.7003e-5) +zircaloy.add_nuclide(sn118, 1.1669e-4) +zircaloy.add_nuclide(sn119, 4.1387e-5) +zircaloy.add_nuclide(sn120, 1.5697e-4) +zircaloy.add_nuclide(sn122, 2.2308e-5) +zircaloy.add_nuclide(sn124, 2.7897e-5) borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm') -borated_water.setDensity('g/cm3', 0.740582) -borated_water.addNuclide(b10, 8.0042e-6) -borated_water.addNuclide(b11, 3.2218e-5) -borated_water.addNuclide(h1, 4.9457e-2) -borated_water.addNuclide(h2, 7.4196e-6) -borated_water.addNuclide(o16, 2.4672e-2) -borated_water.addNuclide(o17, 6.0099e-5) -borated_water.addSAlphaBeta('HH2O', '71t') +borated_water.set_density('g/cm3', 0.740582) +borated_water.add_nuclide(b10, 8.0042e-6) +borated_water.add_nuclide(b11, 3.2218e-5) +borated_water.add_nuclide(h1, 4.9457e-2) +borated_water.add_nuclide(h2, 7.4196e-6) +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() -materials_file.setDefaultXS('71c') -materials_file.addMaterials([uo2, helium, zircaloy, borated_water]) -materials_file.exportToXML() +materials_file.set_default_xs('71c') +materials_file.add_materials([uo2, helium, zircaloy, borated_water]) +materials_file.export_to_xml() ############################################################################### @@ -121,10 +121,10 @@ right = openmc.XPlane(surface_id=5, x0=0.62992, name='right') bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom') top = openmc.YPlane(surface_id=7, y0=0.62992, name='top') -left.setBoundaryType('reflective') -right.setBoundaryType('reflective') -top.setBoundaryType('reflective') -bottom.setBoundaryType('reflective') +left.set_boundary_type('reflective') +right.set_boundary_type('reflective') +top.set_boundary_type('reflective') +bottom.set_boundary_type('reflective') # Instantiate Cells fuel = openmc.Cell(cell_id=1, name='cell 1') @@ -133,37 +133,37 @@ clad = openmc.Cell(cell_id=3, name='cell 3') water = openmc.Cell(cell_id=4, name='cell 4') # Register Surfaces with Cells -fuel.addSurface(fuel_or, halfspace=-1) -gap.addSurface(fuel_or, halfspace=+1) -gap.addSurface(clad_ir, halfspace=-1) -clad.addSurface(clad_ir, halfspace=+1) -clad.addSurface(clad_or, halfspace=-1) -water.addSurface(clad_or, halfspace=+1) -water.addSurface(left, halfspace=+1) -water.addSurface(right, halfspace=-1) -water.addSurface(bottom, halfspace=+1) -water.addSurface(top, halfspace=-1) +fuel.add_surface(fuel_or, halfspace=-1) +gap.add_surface(fuel_or, halfspace=+1) +gap.add_surface(clad_ir, halfspace=-1) +clad.add_surface(clad_ir, halfspace=+1) +clad.add_surface(clad_or, halfspace=-1) +water.add_surface(clad_or, halfspace=+1) +water.add_surface(left, halfspace=+1) +water.add_surface(right, halfspace=-1) +water.add_surface(bottom, halfspace=+1) +water.add_surface(top, halfspace=-1) # Register Materials with Cells -fuel.setFill(uo2) -gap.setFill(helium) -clad.setFill(zircaloy) -water.setFill(borated_water) +fuel.set_fill(uo2) +gap.set_fill(helium) +clad.set_fill(zircaloy) +water.set_fill(borated_water) # Instantiate Universe root = openmc.Universe(universe_id=0, name='root universe') # Register Cells with Universe -root.addCells([fuel, gap, clad, water]) +root.add_cells([fuel, gap, clad, water]) # Instantiate a Geometry and register the root Universe geometry = openmc.Geometry() -geometry.setRootUniverse(root) +geometry.set_root_universe(root) # Instantiate a GeometryFile, register Geometry, and export to XML geometry_file = openmc.GeometryFile() -geometry_file.setGeometry(geometry) -geometry_file.exportToXML() +geometry_file.set_geometry(geometry) +geometry_file.export_to_xml() ############################################################################### @@ -172,15 +172,15 @@ geometry_file.exportToXML() # Instantiate a SettingsFile, set all runtime parameters, and export to XML settings_file = openmc.SettingsFile() -settings_file.setBatches(batches) -settings_file.setInactive(inactive) -settings_file.setParticles(particles) -settings_file.setSourceSpace('box', [-0.62992, -0.62992, -1, \ +settings_file.set_batches(batches) +settings_file.set_inactive(inactive) +settings_file.set_particles(particles) +settings_file.set_source_space('box', [-0.62992, -0.62992, -1, \ 0.62992, 0.62992, 1]) -settings_file.setEntropyLowerLeft([-0.39218, -0.39218, -1.e50]) -settings_file.setEntropyUpperRight([0.39218, 0.39218, 1.e50]) -settings_file.setEntropyDimension([10, 10, 1]) -settings_file.exportToXML() +settings_file.set_entropy_lower_left([-0.39218, -0.39218, -1.e50]) +settings_file.set_entropy_upper_right([0.39218, 0.39218, 1.e50]) +settings_file.set_entropy_dimension([10, 10, 1]) +settings_file.export_to_xml() ############################################################################### @@ -189,26 +189,26 @@ settings_file.exportToXML() # Instantiate a tally mesh mesh = openmc.Mesh(mesh_id=1) -mesh.setType('rectangular') -mesh.setDimension([100, 100, 1]) -mesh.setLowerLeft([-0.62992, -0.62992, -1.e50]) -mesh.setUpperRight([0.62992, 0.62992, 1.e50]) +mesh.set_type('rectangular') +mesh.set_dimension([100, 100, 1]) +mesh.set_lower_left([-0.62992, -0.62992, -1.e50]) +mesh.set_upper_right([0.62992, 0.62992, 1.e50]) # Instantiate some tally Filters energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.]) mesh_filter = openmc.Filter() -mesh_filter.setMesh(mesh) +mesh_filter.set_mesh(mesh) # Instantiate the Tally tally = openmc.Tally(tally_id=1) -tally.addFilter(energy_filter) -tally.addFilter(mesh_filter) -tally.addScore('flux') -tally.addScore('fission') -tally.addScore('nu-fission') +tally.add_filter(energy_filter) +tally.add_filter(mesh_filter) +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() -tallies_file.addMesh(mesh) -tallies_file.addTally(tally) -tallies_file.exportToXML() +tallies_file.add_mesh(mesh) +tallies_file.add_tally(tally) +tallies_file.export_to_xml() diff --git a/examples/python/reflective/build-xml.py b/examples/python/reflective/build-xml.py index ef54e8b2e..4e09eeb97 100644 --- a/examples/python/reflective/build-xml.py +++ b/examples/python/reflective/build-xml.py @@ -20,14 +20,14 @@ u235 = openmc.Nuclide('U-235') # Instantiate a Material and register the Nuclide fuel = openmc.Material(material_id=1, name='fuel') -fuel.setDensity('g/cc', 4.5) -fuel.addNuclide(u235, 1.) +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() -materials_file.setDefaultXS('71c') -materials_file.addMaterial(fuel) -materials_file.exportToXML() +materials_file.set_default_xs('71c') +materials_file.add_material(fuel) +materials_file.export_to_xml() ############################################################################### @@ -42,41 +42,41 @@ surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4') surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5') surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6') -surf1.setBoundaryType('vacuum') -surf2.setBoundaryType('vacuum') -surf3.setBoundaryType('reflective') -surf4.setBoundaryType('reflective') -surf5.setBoundaryType('reflective') -surf6.setBoundaryType('reflective') +surf1.set_boundary_type('vacuum') +surf2.set_boundary_type('vacuum') +surf3.set_boundary_type('reflective') +surf4.set_boundary_type('reflective') +surf5.set_boundary_type('reflective') +surf6.set_boundary_type('reflective') # Instantiate Cell cell = openmc.Cell(cell_id=1, name='cell 1') # Register Surfaces with Cell -cell.addSurface(surface=surf1, halfspace=+1) -cell.addSurface(surface=surf2, halfspace=-1) -cell.addSurface(surface=surf3, halfspace=+1) -cell.addSurface(surface=surf4, halfspace=-1) -cell.addSurface(surface=surf5, halfspace=+1) -cell.addSurface(surface=surf6, halfspace=-1) +cell.add_surface(surface=surf1, halfspace=+1) +cell.add_surface(surface=surf2, halfspace=-1) +cell.add_surface(surface=surf3, halfspace=+1) +cell.add_surface(surface=surf4, halfspace=-1) +cell.add_surface(surface=surf5, halfspace=+1) +cell.add_surface(surface=surf6, halfspace=-1) # Register Material with Cell -cell.setFill(fuel) +cell.set_fill(fuel) # Instantiate Universes root = openmc.Universe(universe_id=0, name='root universe') # Register Cell with Universe -root.addCell(cell) +root.add_cell(cell) # Instantiate a Geometry and register the root Universe geometry = openmc.Geometry() -geometry.setRootUniverse(root) +geometry.set_root_universe(root) # Instantiate a GeometryFile, register Geometry, and export to XML geometry_file = openmc.GeometryFile() -geometry_file.setGeometry(geometry) -geometry_file.exportToXML() +geometry_file.set_geometry(geometry) +geometry_file.export_to_xml() ############################################################################### @@ -85,8 +85,8 @@ geometry_file.exportToXML() # Instantiate a SettingsFile, set all runtime parameters, and export to XML settings_file = openmc.SettingsFile() -settings_file.setBatches(batches) -settings_file.setInactive(inactive) -settings_file.setParticles(particles) -settings_file.setSourceSpace('box', [-1, -1, -1, 1, 1, 1]) -settings_file.exportToXML() +settings_file.set_batches(batches) +settings_file.set_inactive(inactive) +settings_file.set_particles(particles) +settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1]) +settings_file.export_to_xml() diff --git a/examples/python/reflective/geometry.xml b/examples/python/reflective/geometry.xml deleted file mode 100644 index 85ee37f77..000000000 --- a/examples/python/reflective/geometry.xml +++ /dev/null @@ -1,17 +0,0 @@ - - - - - - - - - - - - - - - - - diff --git a/examples/python/reflective/materials.xml b/examples/python/reflective/materials.xml deleted file mode 100644 index 1c6a7fa54..000000000 --- a/examples/python/reflective/materials.xml +++ /dev/null @@ -1,9 +0,0 @@ - - - 71c - - - - - - diff --git a/examples/python/reflective/settings.xml b/examples/python/reflective/settings.xml deleted file mode 100644 index db21c9066..000000000 --- a/examples/python/reflective/settings.xml +++ /dev/null @@ -1,13 +0,0 @@ - - - - 10000 - 500 - 10 - - - - -1 -1 -1 1 1 1 - - -