mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-26 13:15:39 -04:00
Updated sample input scripts for Python API to conform to new API specs
This commit is contained in:
Will Boyd
parent
2767d3f53f
commit
3bb1f8faf7
8 changed files with 291 additions and 330 deletions
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@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235')
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# Instantiate some Materials and register the appropriate Nuclides
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moderator = openmc.Material(material_id=41, name='moderator')
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moderator.setDensity('g/cc', 1.0)
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moderator.addNuclide(h1, 2.)
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moderator.addNuclide(o16, 1.)
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moderator.addSAlphaBeta('HH2O', '71t')
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moderator.set_density('g/cc', 1.0)
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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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fuel = openmc.Material(material_id=40, name='fuel')
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fuel.setDensity('g/cc', 4.5)
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fuel.addNuclide(u235, 1.)
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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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materials_file.setDefaultXS('71c')
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materials_file.addMaterials([moderator, fuel])
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materials_file.exportToXML()
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materials_file.set_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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@ -46,7 +46,7 @@ materials_file.exportToXML()
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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.setBoundaryType('vacuum')
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surf3.set_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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@ -55,34 +55,34 @@ cell3 = openmc.Cell(cell_id=101, name='cell 3')
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cell4 = openmc.Cell(cell_id=2, name='cell 4')
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# Register Surfaces with Cells
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cell1.addSurface(surface=surf2, halfspace=-1)
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cell2.addSurface(surface=surf1, halfspace=-1)
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cell3.addSurface(surface=surf1, halfspace=+1)
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cell4.addSurface(surface=surf2, halfspace=+1)
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cell4.addSurface(surface=surf3, halfspace=-1)
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cell1.add_surface(surface=surf2, halfspace=-1)
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cell2.add_surface(surface=surf1, halfspace=-1)
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cell3.add_surface(surface=surf1, halfspace=+1)
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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.setFill(fuel)
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cell3.setFill(moderator)
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cell4.setFill(moderator)
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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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# 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.setFill(universe1)
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cell1.set_fill(universe1)
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# Register Cells with Universes
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universe1.addCells([cell2, cell3])
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root.addCells([cell1, cell4])
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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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geometry = openmc.Geometry()
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geometry.setRootUniverse(root)
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geometry.set_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.setGeometry(geometry)
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geometry_file.exportToXML()
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geometry_file.set_geometry(geometry)
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geometry_file.export_to_xml()
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###############################################################################
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@ -91,11 +91,11 @@ geometry_file.exportToXML()
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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.setBatches(batches)
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settings_file.setInactive(inactive)
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settings_file.setParticles(particles)
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settings_file.setSourceSpace('box', [-4, -4, -4, 4, 4, 4])
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settings_file.exportToXML()
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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.set_source_space('box', [-4, -4, -4, 4, 4, 4])
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settings_file.export_to_xml()
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###############################################################################
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@ -109,33 +109,33 @@ energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.])
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# Instantiate the first Tally
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first_tally = openmc.Tally(tally_id=1, label='first tally')
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first_tally.addFilter(cell_filter)
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first_tally.add_filter(cell_filter)
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scores = ['total', 'scatter', 'nu-scatter', \
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'absorption', 'fission', 'nu-fission']
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for score in scores:
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first_tally.addScore(score)
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first_tally.add_score(score)
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# Instantiate the second Tally
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second_tally = openmc.Tally(tally_id=2, label='second tally')
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second_tally.addFilter(cell_filter)
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second_tally.addFilter(energy_filter)
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second_tally.add_filter(cell_filter)
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second_tally.add_filter(energy_filter)
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scores = ['total', 'scatter', 'nu-scatter', \
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'absorption', 'fission', 'nu-fission']
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for score in scores:
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second_tally.addScore(score)
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second_tally.add_score(score)
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# Instantiate the third Tally
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third_tally = openmc.Tally(tally_id=3, label='third tally')
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third_tally.addFilter(cell_filter)
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third_tally.addFilter(energy_filter)
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third_tally.addFilter(energyout_filter)
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third_tally.add_filter(cell_filter)
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third_tally.add_filter(energy_filter)
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third_tally.add_filter(energyout_filter)
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scores = ['scatter', 'nu-scatter', 'nu-fission']
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for score in scores:
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third_tally.addScore(score)
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third_tally.add_score(score)
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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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tallies_file.addTally(first_tally)
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tallies_file.addTally(second_tally)
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tallies_file.addTally(third_tally)
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tallies_file.exportToXML()
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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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tallies_file.export_to_xml()
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@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235')
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# Instantiate some Materials and register the appropriate Nuclides
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fuel = openmc.Material(material_id=1, name='fuel')
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fuel.setDensity('g/cc', 4.5)
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fuel.addNuclide(u235, 1.)
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fuel.set_density('g/cc', 4.5)
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fuel.add_nuclide(u235, 1.)
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moderator = openmc.Material(material_id=2, name='moderator')
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moderator.setDensity('g/cc', 1.0)
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moderator.addNuclide(h1, 2.)
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moderator.addNuclide(o16, 1.)
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moderator.addSAlphaBeta('HH2O', '71t')
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moderator.set_density('g/cc', 1.0)
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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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materials_file.setDefaultXS('71c')
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materials_file.addMaterials([moderator, fuel])
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materials_file.exportToXML()
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materials_file.set_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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@ -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.setBoundaryType('vacuum')
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right.setBoundaryType('vacuum')
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top.setBoundaryType('vacuum')
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bottom.setBoundaryType('vacuum')
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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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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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@ -67,28 +67,28 @@ cell7 = openmc.Cell(cell_id=301, name='cell 7')
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cell8 = openmc.Cell(cell_id=302, name='cell 8')
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# Register Surfaces with Cells
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cell1.addSurface(left, halfspace=+1)
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cell1.addSurface(right, halfspace=-1)
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cell1.addSurface(bottom, halfspace=+1)
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cell1.addSurface(top, halfspace=-1)
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cell2.addSurface(left, halfspace=+1)
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cell2.addSurface(right, halfspace=-1)
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cell2.addSurface(bottom, halfspace=+1)
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cell2.addSurface(top, halfspace=-1)
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cell3.addSurface(fuel1, halfspace=-1)
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cell4.addSurface(fuel1, halfspace=+1)
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cell5.addSurface(fuel2, halfspace=-1)
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cell6.addSurface(fuel2, halfspace=+1)
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cell7.addSurface(fuel3, halfspace=-1)
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cell8.addSurface(fuel3, halfspace=+1)
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cell1.add_surface(left, halfspace=+1)
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cell1.add_surface(right, halfspace=-1)
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cell1.add_surface(bottom, halfspace=+1)
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cell1.add_surface(top, halfspace=-1)
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cell2.add_surface(left, halfspace=+1)
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cell2.add_surface(right, halfspace=-1)
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cell2.add_surface(bottom, halfspace=+1)
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cell2.add_surface(top, halfspace=-1)
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cell3.add_surface(fuel1, halfspace=-1)
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cell4.add_surface(fuel1, halfspace=+1)
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cell5.add_surface(fuel2, halfspace=-1)
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cell6.add_surface(fuel2, halfspace=+1)
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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.setFill(fuel)
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cell4.setFill(moderator)
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cell5.setFill(fuel)
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cell6.setFill(moderator)
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cell7.setFill(fuel)
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cell8.setFill(moderator)
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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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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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@ -98,39 +98,39 @@ univ4 = openmc.Universe(universe_id=5)
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cells with Universe
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univ1.addCells([cell3, cell4])
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univ2.addCells([cell5, cell6])
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univ3.addCells([cell7, cell8])
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root.addCell(cell1)
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univ4.addCell(cell2)
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univ1.add_cells([cell3, cell4])
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univ2.add_cells([cell5, cell6])
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univ3.add_cells([cell7, cell8])
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root.add_cell(cell1)
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univ4.add_cell(cell2)
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# Instantiate nested Lattices
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lattice1 = openmc.Lattice(lattice_id=4, name='4x4 assembly')
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lattice1.setDimension([2, 2])
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lattice1.setLowerLeft([-1., -1.])
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lattice1.setWidth([1., 1.])
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lattice1.setUniverses([[univ1, univ2],
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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_width([1., 1.])
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lattice1.set_universes([[univ1, univ2],
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[univ2, univ3]])
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lattice2 = openmc.Lattice(lattice_id=6, name='4x4 core')
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lattice2.setDimension([2, 2])
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lattice2.setLowerLeft([-2., -2.])
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lattice2.setWidth([2., 2.])
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lattice2.setUniverses([[univ4, univ4],
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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_width([2., 2.])
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lattice2.set_universes([[univ4, univ4],
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[univ4, univ4]])
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# Fill Cell with the Lattice
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cell1.setFill(lattice2)
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cell2.setFill(lattice1)
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cell1.set_fill(lattice2)
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cell2.set_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.setRootUniverse(root)
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geometry.set_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.setGeometry(geometry)
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geometry_file.exportToXML()
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geometry_file.set_geometry(geometry)
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geometry_file.export_to_xml()
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###############################################################################
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@ -139,11 +139,11 @@ geometry_file.exportToXML()
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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.setBatches(batches)
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settings_file.setInactive(inactive)
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settings_file.setParticles(particles)
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settings_file.setSourceSpace('box', [-1, -1, -1, 1, 1, 1])
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settings_file.exportToXML()
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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.set_source_space('box', [-1, -1, -1, 1, 1, 1])
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settings_file.export_to_xml()
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###############################################################################
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@ -151,15 +151,15 @@ settings_file.exportToXML()
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###############################################################################
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plot = openmc.Plot(plot_id=1)
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plot.setOrigin([0, 0, 0])
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plot.setWidth([4, 4])
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plot.setPixels([400, 400])
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plot.setColor('mat')
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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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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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plot_file.addPlot(plot)
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plot_file.exportToXML()
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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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@ -168,22 +168,22 @@ plot_file.exportToXML()
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# Instantiate a tally mesh
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mesh = openmc.Mesh(mesh_id=1)
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mesh.setType('rectangular')
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mesh.setDimension([4, 4])
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mesh.setLowerLeft([-2, -2])
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mesh.setWidth([1, 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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# Instantiate tally Filter
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mesh_filter = openmc.Filter()
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mesh_filter.setMesh(mesh)
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mesh_filter.set_mesh(mesh)
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# Instantiate the Tally
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tally = openmc.Tally(tally_id=1)
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tally.addFilter(mesh_filter)
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tally.addScore('total')
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tally.add_filter(mesh_filter)
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tally.add_score('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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tallies_file.addMesh(mesh)
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tallies_file.addTally(tally)
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tallies_file.exportToXML()
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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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@ -22,20 +22,20 @@ u235 = openmc.Nuclide('U-235')
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# Instantiate some Materials and register the appropriate Nuclides
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fuel = openmc.Material(material_id=1, name='fuel')
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fuel.setDensity('g/cc', 4.5)
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fuel.addNuclide(u235, 1.)
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fuel.set_density('g/cc', 4.5)
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fuel.add_nuclide(u235, 1.)
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moderator = openmc.Material(material_id=2, name='moderator')
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moderator.setDensity('g/cc', 1.0)
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moderator.addNuclide(h1, 2.)
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moderator.addNuclide(o16, 1.)
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moderator.addSAlphaBeta('HH2O', '71t')
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moderator.set_density('g/cc', 1.0)
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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')
|
||||
|
||||
# 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()
|
||||
|
|
|
|||
|
|
@ -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()
|
||||
|
|
|
|||
|
|
@ -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()
|
||||
|
|
|
|||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<geometry>
|
||||
<!--cell 1-->
|
||||
<cell id="1" material="1" surfaces="1 -2 3 -4 5 -6" universe="0" />
|
||||
<!--surf 1-->
|
||||
<surface boundary="vacuum" coeffs="-1" id="1" type="x-plane" />
|
||||
<!--surf 2-->
|
||||
<surface boundary="vacuum" coeffs="1" id="2" type="x-plane" />
|
||||
<!--surf 3-->
|
||||
<surface boundary="reflective" coeffs="-1" id="3" type="y-plane" />
|
||||
<!--surf 4-->
|
||||
<surface boundary="reflective" coeffs="1" id="4" type="y-plane" />
|
||||
<!--surf 5-->
|
||||
<surface boundary="reflective" coeffs="-1" id="5" type="z-plane" />
|
||||
<!--surf 6-->
|
||||
<surface boundary="reflective" coeffs="1" id="6" type="z-plane" />
|
||||
</geometry>
|
||||
|
|
@ -1,9 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<materials>
|
||||
<default_xs>71c</default_xs>
|
||||
<!--fuel-->
|
||||
<material id="1">
|
||||
<density units="g/cc" value="4.5" />
|
||||
<nuclide ao="1.0" name="U-235" />
|
||||
</material>
|
||||
</materials>
|
||||
|
|
@ -1,13 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<settings>
|
||||
<eigenvalue>
|
||||
<particles>10000</particles>
|
||||
<batches>500</batches>
|
||||
<inactive>10</inactive>
|
||||
</eigenvalue>
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>-1 -1 -1 1 1 1</parameters>
|
||||
</space>
|
||||
</source>
|
||||
</settings>
|
||||
Loading…
Add table
Add a link
Reference in a new issue