mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-25 04:25:29 -04:00
189 lines
6 KiB
Python
189 lines
6 KiB
Python
import openmc
|
|
|
|
|
|
###############################################################################
|
|
# Simulation Input File Parameters
|
|
###############################################################################
|
|
|
|
# OpenMC simulation parameters
|
|
batches = 20
|
|
inactive = 10
|
|
particles = 10000
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC materials.xml File
|
|
###############################################################################
|
|
|
|
# Instantiate some Nuclides
|
|
h1 = openmc.Nuclide('H-1')
|
|
o16 = openmc.Nuclide('O-16')
|
|
u235 = openmc.Nuclide('U-235')
|
|
|
|
# Instantiate some Materials and register the appropriate Nuclides
|
|
fuel = openmc.Material(material_id=1, name='fuel')
|
|
fuel.set_density('g/cc', 4.5)
|
|
fuel.add_nuclide(u235, 1.)
|
|
|
|
moderator = openmc.Material(material_id=2, name='moderator')
|
|
moderator.set_density('g/cc', 1.0)
|
|
moderator.add_nuclide(h1, 2.)
|
|
moderator.add_nuclide(o16, 1.)
|
|
moderator.add_s_alpha_beta('HH2O', '71t')
|
|
|
|
# Instantiate a MaterialsFile, register all Materials, and export to XML
|
|
materials_file = openmc.MaterialsFile()
|
|
materials_file.default_xs = '71c'
|
|
materials_file.add_materials([moderator, fuel])
|
|
materials_file.export_to_xml()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC geometry.xml File
|
|
###############################################################################
|
|
|
|
# Instantiate Surfaces
|
|
left = openmc.XPlane(surface_id=1, x0=-2, name='left')
|
|
right = openmc.XPlane(surface_id=2, x0=2, name='right')
|
|
bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
|
|
top = openmc.YPlane(surface_id=4, y0=2, name='top')
|
|
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.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')
|
|
cell2 = openmc.Cell(cell_id=2, name='Cell 2')
|
|
cell3 = openmc.Cell(cell_id=101, name='cell 3')
|
|
cell4 = openmc.Cell(cell_id=102, name='cell 4')
|
|
cell5 = openmc.Cell(cell_id=201, name='cell 5')
|
|
cell6 = openmc.Cell(cell_id=202, name='cell 6')
|
|
cell7 = openmc.Cell(cell_id=301, name='cell 7')
|
|
cell8 = openmc.Cell(cell_id=302, name='cell 8')
|
|
|
|
# Register Surfaces with Cells
|
|
cell1.add_surface(left, halfspace=+1)
|
|
cell1.add_surface(right, halfspace=-1)
|
|
cell1.add_surface(bottom, halfspace=+1)
|
|
cell1.add_surface(top, halfspace=-1)
|
|
cell2.add_surface(left, halfspace=+1)
|
|
cell2.add_surface(right, halfspace=-1)
|
|
cell2.add_surface(bottom, halfspace=+1)
|
|
cell2.add_surface(top, halfspace=-1)
|
|
cell3.add_surface(fuel1, halfspace=-1)
|
|
cell4.add_surface(fuel1, halfspace=+1)
|
|
cell5.add_surface(fuel2, halfspace=-1)
|
|
cell6.add_surface(fuel2, halfspace=+1)
|
|
cell7.add_surface(fuel3, halfspace=-1)
|
|
cell8.add_surface(fuel3, halfspace=+1)
|
|
|
|
# Register Materials with Cells
|
|
cell3.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)
|
|
univ2 = openmc.Universe(universe_id=2)
|
|
univ3 = openmc.Universe(universe_id=3)
|
|
univ4 = openmc.Universe(universe_id=5)
|
|
root = openmc.Universe(universe_id=0, name='root universe')
|
|
|
|
# Register Cells with Universe
|
|
univ1.add_cells([cell3, cell4])
|
|
univ2.add_cells([cell5, cell6])
|
|
univ3.add_cells([cell7, cell8])
|
|
root.add_cell(cell1)
|
|
univ4.add_cell(cell2)
|
|
|
|
# Instantiate nested Lattices
|
|
lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
|
|
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.dimension = [2, 2]
|
|
lattice2.lower_left = [-2., -2.]
|
|
lattice2.pitch = [2., 2.]
|
|
lattice2.universes = [[univ4, univ4],
|
|
[univ4, univ4]]
|
|
|
|
# Fill Cell with the Lattice
|
|
cell1.fill = lattice2
|
|
cell2.fill = lattice1
|
|
|
|
# Instantiate a Geometry and register the root Universe
|
|
geometry = openmc.Geometry()
|
|
geometry.root_universe = root
|
|
|
|
# Instantiate a GeometryFile, register Geometry, and export to XML
|
|
geometry_file = openmc.GeometryFile()
|
|
geometry_file.geometry = geometry
|
|
geometry_file.export_to_xml()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC settings.xml File
|
|
###############################################################################
|
|
|
|
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
|
|
settings_file = openmc.SettingsFile()
|
|
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()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC plots.xml File
|
|
###############################################################################
|
|
|
|
plot = openmc.Plot(plot_id=1)
|
|
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()
|
|
plot_file.add_plot(plot)
|
|
plot_file.export_to_xml()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC tallies.xml File
|
|
###############################################################################
|
|
|
|
# Instantiate a tally mesh
|
|
mesh = openmc.Mesh(mesh_id=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.mesh = mesh
|
|
|
|
# Instantiate the Tally
|
|
tally = openmc.Tally(tally_id=1)
|
|
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.add_mesh(mesh)
|
|
tallies_file.add_tally(tally)
|
|
tallies_file.export_to_xml()
|