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Geometric intrinsics now use property setter/getter decorators

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This commit is contained in:
Will Boyd 2015-04-22 00:50:03 -05:00
parent 9cada4df0e
commit 5dac75297d
14 changed files with 1014 additions and 466 deletions
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74
examples/python/lattice/nested/build-xml.py
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@ -33,7 +33,7 @@ moderator.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([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.set_boundary_type('vacuum')
right.set_boundary_type('vacuum')
top.set_boundary_type('vacuum')
bottom.set_boundary_type('vacuum')
left.boundary_type = 'vacuum'
right.boundary_type = 'vacuum'
top.boundary_type = 'vacuum'
bottom.boundary_type = 'vacuum'
# Instantiate Cells
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
@ -83,12 +83,12 @@ cell7.add_surface(fuel3, halfspace=-1)
cell8.add_surface(fuel3, halfspace=+1)
# Register Materials with Cells
cell3.set_fill(fuel)
cell4.set_fill(moderator)
cell5.set_fill(fuel)
cell6.set_fill(moderator)
cell7.set_fill(fuel)
cell8.set_fill(moderator)
cell3.fill = fuel
cell4.fill = moderator
cell5.fill = fuel
cell6.fill = moderator
cell7.fill = fuel
cell8.fill = moderator
# Instantiate Universe
univ1 = openmc.Universe(universe_id=1)
@ -106,30 +106,30 @@ univ4.add_cell(cell2)
# Instantiate nested Lattices
lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
lattice1.set_dimension([2, 2])
lattice1.set_lower_left([-1., -1.])
lattice1.set_pitch([1., 1.])
lattice1.set_universes([[univ1, univ2],
[univ2, univ3]])
lattice1.dimension = [2, 2]
lattice1.lower_left = [-1., -1.]
lattice1.pitch = [1., 1.]
lattice1.universes = [[univ1, univ2],
[univ2, univ3]]
lattice2 = openmc.RectLattice(lattice_id=6, name='4x4 core')
lattice2.set_dimension([2, 2])
lattice2.set_lower_left([-2., -2.])
lattice2.set_pitch([2., 2.])
lattice2.set_universes([[univ4, univ4],
[univ4, univ4]])
lattice2.dimension = [2, 2]
lattice2.lower_left = [-2., -2.]
lattice2.pitch = [2., 2.]
lattice2.universes = [[univ4, univ4],
[univ4, univ4]]
# Fill Cell with the Lattice
cell1.set_fill(lattice2)
cell2.set_fill(lattice1)
cell1.fill = lattice2
cell2.fill = lattice1
# 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()
@ -139,9 +139,9 @@ 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()
@ -151,10 +151,10 @@ 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()
@ -168,14 +168,14 @@ 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)