mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-22 15:05:28 -04:00
166 lines
5.3 KiB
Python
166 lines
5.3 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 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_element('H', 2.)
|
|
moderator.add_element('O', 1.)
|
|
moderator.add_s_alpha_beta('c_H_in_H2O')
|
|
|
|
# Instantiate a Materials collection and export to XML
|
|
materials_file = openmc.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=101, name='cell 2')
|
|
cell3 = openmc.Cell(cell_id=102, name='cell 3')
|
|
cell4 = openmc.Cell(cell_id=201, name='cell 4')
|
|
cell5 = openmc.Cell(cell_id=202, name='cell 5')
|
|
cell6 = openmc.Cell(cell_id=301, name='cell 6')
|
|
cell7 = openmc.Cell(cell_id=302, name='cell 7')
|
|
|
|
# Use surface half-spaces to define regions
|
|
cell1.region = +left & -right & +bottom & -top
|
|
cell2.region = -fuel1
|
|
cell3.region = +fuel1
|
|
cell4.region = -fuel2
|
|
cell5.region = +fuel2
|
|
cell6.region = -fuel3
|
|
cell7.region = +fuel3
|
|
|
|
# Register Materials with Cells
|
|
cell2.fill = fuel
|
|
cell3.fill = moderator
|
|
cell4.fill = fuel
|
|
cell5.fill = moderator
|
|
cell6.fill = fuel
|
|
cell7.fill = moderator
|
|
|
|
# Instantiate Universe
|
|
univ1 = openmc.Universe(universe_id=1)
|
|
univ2 = openmc.Universe(universe_id=2)
|
|
univ3 = openmc.Universe(universe_id=3)
|
|
root = openmc.Universe(universe_id=0, name='root universe')
|
|
|
|
# Register Cells with Universe
|
|
univ1.add_cells([cell2, cell3])
|
|
univ2.add_cells([cell4, cell5])
|
|
univ3.add_cells([cell6, cell7])
|
|
root.add_cell(cell1)
|
|
|
|
# Instantiate a Lattice
|
|
lattice = openmc.RectLattice(lattice_id=5)
|
|
lattice.lower_left = [-2., -2.]
|
|
lattice.pitch = [1., 1.]
|
|
lattice.universes = [[univ1, univ2, univ1, univ2],
|
|
[univ2, univ3, univ2, univ3],
|
|
[univ1, univ2, univ1, univ2],
|
|
[univ2, univ3, univ2, univ3]]
|
|
|
|
# Fill Cell with the Lattice
|
|
cell1.fill = lattice
|
|
|
|
# Instantiate a Geometry, register the root Universe, and export to XML
|
|
geometry = openmc.Geometry(root)
|
|
geometry.export_to_xml()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC settings.xml file
|
|
###############################################################################
|
|
|
|
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
|
settings_file = openmc.Settings()
|
|
settings_file.batches = batches
|
|
settings_file.inactive = inactive
|
|
settings_file.particles = particles
|
|
|
|
# Create an initial uniform spatial source distribution over fissionable zones
|
|
bounds = [-1, -1, -1, 1, 1, 1]
|
|
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
|
settings_file.source = openmc.source.Source(space=uniform_dist)
|
|
|
|
settings_file.trigger_active = True
|
|
settings_file.trigger_max_batches = 100
|
|
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_by = 'material'
|
|
|
|
# Instantiate a Plots collection and export to XML
|
|
plot_file = openmc.Plots([plot])
|
|
plot_file.export_to_xml()
|
|
|
|
|
|
###############################################################################
|
|
# Exporting to OpenMC tallies.xml file
|
|
###############################################################################
|
|
|
|
# Instantiate a tally mesh
|
|
mesh = openmc.RegularMesh(mesh_id=1)
|
|
mesh.dimension = [4, 4]
|
|
mesh.lower_left = [-2, -2]
|
|
mesh.width = [1, 1]
|
|
|
|
# Instantiate tally Filter
|
|
mesh_filter = openmc.MeshFilter(mesh)
|
|
|
|
# Instantiate tally Trigger
|
|
trigger = openmc.Trigger(trigger_type='rel_err', threshold=1E-2)
|
|
trigger.scores = ['all']
|
|
|
|
# Instantiate the Tally
|
|
tally = openmc.Tally(tally_id=1)
|
|
tally.filters = [mesh_filter]
|
|
tally.scores = ['total']
|
|
tally.triggers = [trigger]
|
|
|
|
# Instantiate a Tallies collection and export to XML
|
|
tallies_file = openmc.Tallies([tally])
|
|
tallies_file.export_to_xml()
|