mirror of
https://github.com/openmc-dev/openmc.git
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Geometric intrinsics now use property setter/getter decorators
This commit is contained in:
Will Boyd
parent
9cada4df0e
commit
5dac75297d
14 changed files with 1014 additions and 466 deletions
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@ -33,7 +33,7 @@ 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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materials_file.set_default_xs('71c')
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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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@ -46,7 +46,7 @@ materials_file.export_to_xml()
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surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=7, name='surf 1')
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surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=9, name='surf 2')
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surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=11, name='surf 3')
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surf3.set_boundary_type('vacuum')
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surf3.boundary_type = 'vacuum'
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='cell 1')
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@ -62,14 +62,14 @@ cell4.add_surface(surface=surf2, halfspace=+1)
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cell4.add_surface(surface=surf3, halfspace=-1)
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# Register Materials with Cells
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cell2.set_fill(fuel)
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cell3.set_fill(moderator)
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cell4.set_fill(moderator)
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cell2.fill = fuel
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cell3.fill = moderator
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cell4.fill = moderator
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# Instantiate Universes
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universe1 = openmc.Universe(universe_id=37)
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root = openmc.Universe(universe_id=0, name='root universe')
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cell1.set_fill(universe1)
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cell1.fill = universe1
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# Register Cells with Universes
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universe1.add_cells([cell2, cell3])
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@ -77,11 +77,11 @@ root.add_cells([cell1, cell4])
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.set_root_universe(root)
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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.set_geometry(geometry)
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geometry_file.geometry = geometry
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geometry_file.export_to_xml()
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@ -91,9 +91,9 @@ geometry_file.export_to_xml()
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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.set_batches(batches)
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settings_file.set_inactive(inactive)
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settings_file.set_particles(particles)
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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settings_file.set_source_space('box', [-4, -4, -4, 4, 4, 4])
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settings_file.export_to_xml()
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@ -38,7 +38,7 @@ 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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materials_file.set_default_xs('71c')
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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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@ -54,10 +54,10 @@ bottom = openmc.YPlane(surface_id=3, y0=-4, name='bottom')
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top = openmc.YPlane(surface_id=4, y0=4, name='top')
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fuel_surf = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
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left.set_boundary_type('vacuum')
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right.set_boundary_type('vacuum')
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top.set_boundary_type('vacuum')
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bottom.set_boundary_type('vacuum')
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left.boundary_type = 'vacuum'
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right.boundary_type = 'vacuum'
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top.boundary_type = 'vacuum'
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bottom.boundary_type = 'vacuum'
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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@ -78,11 +78,11 @@ cell5.add_surface(fuel_surf, halfspace=-1)
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cell6.add_surface(fuel_surf, halfspace=+1)
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# Register Materials with Cells
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cell2.set_fill(fuel)
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cell3.set_fill(moderator)
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cell4.set_fill(moderator)
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cell5.set_fill(iron)
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cell6.set_fill(moderator)
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cell2.fill = fuel
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cell3.fill = moderator
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cell4.fill = moderator
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cell5.fill = iron
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cell6.fill = moderator
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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@ -98,24 +98,24 @@ root.add_cell(cell1)
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# Instantiate a Lattice
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lattice = openmc.HexLattice(lattice_id=5)
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lattice.set_center([0., 0., 0.])
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lattice.set_pitch([1., 2.])
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lattice.set_universes([
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[ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ],
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[ [univ2] + [univ1]*11, [univ2] + [univ1]*5, [univ1] ],
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[ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ]])
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lattice.set_outer(univ2)
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lattice.center = [0., 0., 0.]
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lattice.pitch = [1., 2.]
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lattice.universes = \
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[ [ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ],
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[ [univ2] + [univ1]*11, [univ2] + [univ1]*5, [univ1] ],
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[ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ] ]
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lattice.outer = univ2
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# Fill Cell with the Lattice
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cell1.set_fill(lattice)
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cell1.fill = lattice
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.set_root_universe(root)
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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.set_geometry(geometry)
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geometry_file.geometry = geometry
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geometry_file.export_to_xml()
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@ -125,9 +125,9 @@ geometry_file.export_to_xml()
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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.set_batches(batches)
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settings_file.set_inactive(inactive)
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settings_file.set_particles(particles)
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
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settings_file.export_to_xml()
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@ -137,19 +137,19 @@ settings_file.export_to_xml()
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###############################################################################
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plot_xy = openmc.Plot(plot_id=1)
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plot_xy.set_filename('plot_xy')
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plot_xy.set_origin([0, 0, 0])
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plot_xy.set_width([6, 6])
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plot_xy.set_pixels([400, 400])
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plot_xy.set_color('mat')
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plot_xy.filename = 'plot_xy'
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plot_xy.origin = [0, 0, 0]
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plot_xy.width = [6, 6]
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plot_xy.pixels = [400, 400]
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plot_xy.color = 'mat'
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plot_yz = openmc.Plot(plot_id=2)
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plot_yz.set_filename('plot_yz')
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plot_yz.set_basis('yz')
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plot_yz.set_origin([0, 0, 0])
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plot_yz.set_width([8, 8])
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plot_yz.set_pixels([400, 400])
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plot_yz.set_color('mat')
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plot_yz.filename = 'plot_yz'
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plot_yz.basis = 'yz'
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plot_yz.origin = [0, 0, 0]
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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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@ -33,7 +33,7 @@ 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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materials_file.set_default_xs('71c')
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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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@ -51,10 +51,10 @@ fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
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fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
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fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
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left.set_boundary_type('vacuum')
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right.set_boundary_type('vacuum')
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top.set_boundary_type('vacuum')
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bottom.set_boundary_type('vacuum')
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left.boundary_type = 'vacuum'
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right.boundary_type = 'vacuum'
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top.boundary_type = 'vacuum'
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bottom.boundary_type = 'vacuum'
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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@ -83,12 +83,12 @@ cell7.add_surface(fuel3, halfspace=-1)
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cell8.add_surface(fuel3, halfspace=+1)
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# Register Materials with Cells
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cell3.set_fill(fuel)
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cell4.set_fill(moderator)
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cell5.set_fill(fuel)
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cell6.set_fill(moderator)
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cell7.set_fill(fuel)
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cell8.set_fill(moderator)
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cell3.fill = fuel
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cell4.fill = moderator
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cell5.fill = fuel
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cell6.fill = moderator
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cell7.fill = fuel
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cell8.fill = moderator
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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@ -106,30 +106,30 @@ univ4.add_cell(cell2)
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# Instantiate nested Lattices
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lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
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lattice1.set_dimension([2, 2])
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lattice1.set_lower_left([-1., -1.])
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lattice1.set_pitch([1., 1.])
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lattice1.set_universes([[univ1, univ2],
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[univ2, univ3]])
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lattice1.dimension = [2, 2]
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lattice1.lower_left = [-1., -1.]
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lattice1.pitch = [1., 1.]
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lattice1.universes = [[univ1, univ2],
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[univ2, univ3]]
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lattice2 = openmc.RectLattice(lattice_id=6, name='4x4 core')
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lattice2.set_dimension([2, 2])
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lattice2.set_lower_left([-2., -2.])
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lattice2.set_pitch([2., 2.])
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lattice2.set_universes([[univ4, univ4],
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[univ4, univ4]])
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lattice2.dimension = [2, 2]
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lattice2.lower_left = [-2., -2.]
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lattice2.pitch = [2., 2.]
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lattice2.universes = [[univ4, univ4],
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[univ4, univ4]]
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# Fill Cell with the Lattice
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cell1.set_fill(lattice2)
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cell2.set_fill(lattice1)
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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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geometry = openmc.Geometry()
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geometry.set_root_universe(root)
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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.set_geometry(geometry)
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geometry_file.geometry = geometry
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geometry_file.export_to_xml()
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@ -139,9 +139,9 @@ geometry_file.export_to_xml()
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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.set_batches(batches)
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settings_file.set_inactive(inactive)
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settings_file.set_particles(particles)
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
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settings_file.export_to_xml()
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@ -151,10 +151,10 @@ settings_file.export_to_xml()
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###############################################################################
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plot = openmc.Plot(plot_id=1)
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plot.set_origin([0, 0, 0])
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plot.set_width([4, 4])
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plot.set_pixels([400, 400])
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plot.set_color('mat')
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plot.origin = [0, 0, 0]
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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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@ -168,14 +168,14 @@ plot_file.export_to_xml()
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# Instantiate a tally mesh
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mesh = openmc.Mesh(mesh_id=1)
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mesh.set_type('rectangular')
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mesh.set_dimension([4, 4])
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mesh.set_lower_left([-2, -2])
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mesh.set_width([1, 1])
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mesh.type = 'rectangular'
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mesh.dimension = [4, 4]
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mesh.lower_left = [-2, -2]
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mesh.width = [1, 1]
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# Instantiate tally Filter
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mesh_filter = openmc.Filter()
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mesh_filter.set_mesh(mesh)
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mesh_filter.mesh = mesh
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# Instantiate the Tally
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tally = openmc.Tally(tally_id=1)
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@ -33,7 +33,7 @@ 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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materials_file.set_default_xs('71c')
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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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@ -51,10 +51,10 @@ fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
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fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
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fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
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left.set_boundary_type('vacuum')
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right.set_boundary_type('vacuum')
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top.set_boundary_type('vacuum')
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bottom.set_boundary_type('vacuum')
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left.boundary_type = 'vacuum'
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right.boundary_type = 'vacuum'
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top.boundary_type = 'vacuum'
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bottom.boundary_type = 'vacuum'
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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@ -78,12 +78,12 @@ cell6.add_surface(fuel3, halfspace=-1)
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cell7.add_surface(fuel3, halfspace=+1)
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# Register Materials with Cells
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cell2.set_fill(fuel)
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cell3.set_fill(moderator)
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cell4.set_fill(fuel)
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cell5.set_fill(moderator)
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cell6.set_fill(fuel)
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cell7.set_fill(moderator)
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cell2.fill = fuel
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cell3.fill = moderator
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cell4.fill = fuel
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cell5.fill = moderator
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cell6.fill = fuel
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cell7.fill = moderator
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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@ -99,24 +99,24 @@ root.add_cell(cell1)
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# Instantiate a Lattice
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lattice = openmc.RectLattice(lattice_id=5)
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lattice.set_dimension([4, 4])
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lattice.set_lower_left([-2., -2.])
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lattice.set_pitch([1., 1.])
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lattice.set_universes([[univ1, univ2, univ1, univ2],
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[univ2, univ3, univ2, univ3],
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[univ1, univ2, univ1, univ2],
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[univ2, univ3, univ2, univ3]])
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lattice.dimension = [4, 4]
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lattice.lower_left = [-2., -2.]
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lattice.pitch = [1., 1.]
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lattice.universes = [[univ1, univ2, univ1, univ2],
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[univ2, univ3, univ2, univ3],
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[univ1, univ2, univ1, univ2],
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[univ2, univ3, univ2, univ3]]
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# Fill Cell with the Lattice
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cell1.set_fill(lattice)
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cell1.fill = lattice
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.set_root_universe(root)
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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.set_geometry(geometry)
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geometry_file.geometry = geometry
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geometry_file.export_to_xml()
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@ -126,9 +126,9 @@ geometry_file.export_to_xml()
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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.set_batches(batches)
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settings_file.set_inactive(inactive)
|
||||
settings_file.set_particles(particles)
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
|
@ -138,10 +138,10 @@ settings_file.export_to_xml()
|
|||
###############################################################################
|
||||
|
||||
plot = openmc.Plot(plot_id=1)
|
||||
plot.set_origin([0, 0, 0])
|
||||
plot.set_width([4, 4])
|
||||
plot.set_pixels([400, 400])
|
||||
plot.set_color('mat')
|
||||
plot.origin = [0, 0, 0]
|
||||
plot.width = [4, 4]
|
||||
plot.pixels = [400, 400]
|
||||
plot.color = 'mat'
|
||||
|
||||
# Instantiate a PlotsFile, add Plot, and export to XML
|
||||
plot_file = openmc.PlotsFile()
|
||||
|
|
@ -155,14 +155,14 @@ plot_file.export_to_xml()
|
|||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.Mesh(mesh_id=1)
|
||||
mesh.set_type('rectangular')
|
||||
mesh.set_dimension([4, 4])
|
||||
mesh.set_lower_left([-2, -2])
|
||||
mesh.set_width([1, 1])
|
||||
mesh.type = 'rectangular'
|
||||
mesh.dimension = [4, 4]
|
||||
mesh.lower_left = [-2, -2]
|
||||
mesh.width = [1, 1]
|
||||
|
||||
# Instantiate tally Filter
|
||||
mesh_filter = openmc.Filter()
|
||||
mesh_filter.set_mesh(mesh)
|
||||
mesh_filter.mesh = mesh
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1)
|
||||
|
|
|
|||
|
|
@ -103,7 +103,7 @@ borated_water.add_s_alpha_beta('HH2O', '71t')
|
|||
|
||||
# Instantiate a MaterialsFile, register all Materials, and export to XML
|
||||
materials_file = openmc.MaterialsFile()
|
||||
materials_file.set_default_xs('71c')
|
||||
materials_file.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.set_boundary_type('reflective')
|
||||
right.set_boundary_type('reflective')
|
||||
top.set_boundary_type('reflective')
|
||||
bottom.set_boundary_type('reflective')
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
|
|
@ -145,10 +145,10 @@ water.add_surface(bottom, halfspace=+1)
|
|||
water.add_surface(top, halfspace=-1)
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.set_fill(uo2)
|
||||
gap.set_fill(helium)
|
||||
clad.set_fill(zircaloy)
|
||||
water.set_fill(borated_water)
|
||||
fuel.fill = uo2
|
||||
gap.fill = helium
|
||||
clad.fill = zircaloy
|
||||
water.fill = borated_water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
|
@ -158,11 +158,11 @@ root.add_cells([fuel, gap, clad, water])
|
|||
|
||||
# Instantiate a Geometry and register the root Universe
|
||||
geometry = openmc.Geometry()
|
||||
geometry.set_root_universe(root)
|
||||
geometry.root_universe = root
|
||||
|
||||
# Instantiate a GeometryFile, register Geometry, and export to XML
|
||||
geometry_file = openmc.GeometryFile()
|
||||
geometry_file.set_geometry(geometry)
|
||||
geometry_file.geometry = geometry
|
||||
geometry_file.export_to_xml()
|
||||
|
||||
|
||||
|
|
@ -172,14 +172,14 @@ geometry_file.export_to_xml()
|
|||
|
||||
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.SettingsFile()
|
||||
settings_file.set_batches(batches)
|
||||
settings_file.set_inactive(inactive)
|
||||
settings_file.set_particles(particles)
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
settings_file.set_source_space('box', [-0.62992, -0.62992, -1, \
|
||||
0.62992, 0.62992, 1])
|
||||
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.entropy_lower_left = [-0.39218, -0.39218, -1.e50]
|
||||
settings_file.entropy_upper_right = [0.39218, 0.39218, 1.e50]
|
||||
settings_file.entropy_dimension = [10, 10, 1]
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
|
|
@ -189,15 +189,15 @@ settings_file.export_to_xml()
|
|||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.Mesh(mesh_id=1)
|
||||
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])
|
||||
mesh.type = 'rectangular'
|
||||
mesh.dimension = [100, 100, 1]
|
||||
mesh.lower_left = [-0.62992, -0.62992, -1.e50]
|
||||
mesh.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.set_mesh(mesh)
|
||||
mesh_filter.mesh = mesh
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1)
|
||||
|
|
|
|||
|
|
@ -25,7 +25,7 @@ fuel.add_nuclide(u235, 1.)
|
|||
|
||||
# Instantiate a MaterialsFile, register Material, and export to XML
|
||||
materials_file = openmc.MaterialsFile()
|
||||
materials_file.set_default_xs('71c')
|
||||
materials_file.default_xs = '71c'
|
||||
materials_file.add_material(fuel)
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
|
@ -42,12 +42,12 @@ 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.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')
|
||||
surf1.boundary_type = 'vacuum'
|
||||
surf2.boundary_type = 'vacuum'
|
||||
surf3.boundary_type = 'reflective'
|
||||
surf4.boundary_type = 'reflective'
|
||||
surf5.boundary_type = 'reflective'
|
||||
surf6.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cell
|
||||
cell = openmc.Cell(cell_id=1, name='cell 1')
|
||||
|
|
@ -61,7 +61,7 @@ cell.add_surface(surface=surf5, halfspace=+1)
|
|||
cell.add_surface(surface=surf6, halfspace=-1)
|
||||
|
||||
# Register Material with Cell
|
||||
cell.set_fill(fuel)
|
||||
cell.fill = fuel
|
||||
|
||||
# Instantiate Universes
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
|
@ -71,11 +71,11 @@ root.add_cell(cell)
|
|||
|
||||
# Instantiate a Geometry and register the root Universe
|
||||
geometry = openmc.Geometry()
|
||||
geometry.set_root_universe(root)
|
||||
geometry.root_universe = root
|
||||
|
||||
# Instantiate a GeometryFile, register Geometry, and export to XML
|
||||
geometry_file = openmc.GeometryFile()
|
||||
geometry_file.set_geometry(geometry)
|
||||
geometry_file.geometry = geometry
|
||||
geometry_file.export_to_xml()
|
||||
|
||||
|
||||
|
|
@ -85,8 +85,8 @@ geometry_file.export_to_xml()
|
|||
|
||||
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.SettingsFile()
|
||||
settings_file.set_batches(batches)
|
||||
settings_file.set_inactive(inactive)
|
||||
settings_file.set_particles(particles)
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
|
||||
settings_file.export_to_xml()
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue