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revise codes related to white boundary condition
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Jiankai Yu
parent
5f394b2816
commit
723153db5c
3 changed files with 0 additions and 148 deletions
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import numpy as np
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import openmc
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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# OpenMC simulation parameters
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batches = 500
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inactive = 100
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particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate a Material and register the Nuclide
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fuel = openmc.Material(material_id=1, name='fuel')
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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 Materials collection and export to XML
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materials_file = openmc.Materials([fuel])
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materials_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate Surfaces
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surf1 = openmc.ZPlane(surface_id=1, z0=0.0, name='surf 1')
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surf2 = openmc.Sphere(surface_id=2, x0=0.0, y0=0.0, z0=0.0, r=0.39218, name='surf 2')
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surf1.boundary_type = 'vacuum'
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surf2.boundary_type = 'white'
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# Instantiate Cell
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cell = openmc.Cell(cell_id=1, name='cell 1')
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# Use surface half-spaces to define region
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cell.region = -surf1 & -surf2
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# Register Material with Cell
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cell.fill = fuel
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# Instantiate Universes
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cell with Universe
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root.add_cell(cell)
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# Instantiate a Geometry, register the root Universe, and export to XML
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geometry = openmc.Geometry(root)
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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml file
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###############################################################################
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# Instantiate a Settings object, set all runtime parameters, and export to XML
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settings_file = openmc.Settings()
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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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# Create an initial uniform spatial source distribution over fissionable zones
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uniform_dist = openmc.stats.Point(xyz=(0, 0, -0.1))
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settings_file.source = openmc.source.Source(space=uniform_dist)
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settings_file.export_to_xml()
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@ -1,75 +0,0 @@
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import numpy as np
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import openmc
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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# OpenMC simulation parameters
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batches = 500
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inactive = 100
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particles = 10000
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###############################################################################
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate a Material and register the Nuclide
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fuel = openmc.Material(material_id=1, name='fuel')
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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 Materials collection and export to XML
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materials_file = openmc.Materials([fuel])
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materials_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate Surfaces
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surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='surf 1')
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surf2 = openmc.ZPlane(surface_id=2, z0=-1, name='surf 2')
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surf3 = openmc.ZPlane(surface_id=3, z0=+1, name='surf 3')
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surf1.boundary_type = 'white'
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surf2.boundary_type = 'vacuum'
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surf3.boundary_type = 'vacuum'
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# Instantiate Cell
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cell = openmc.Cell(cell_id=1, name='cell 1')
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# Use surface half-spaces to define region
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cell.region = -surf1 & +surf2 & -surf3
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# Register Material with Cell
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cell.fill = fuel
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# Instantiate Universes
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cell with Universe
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root.add_cell(cell)
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# Instantiate a Geometry, register the root Universe, and export to XML
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geometry = openmc.Geometry(root)
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geometry.export_to_xml()
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###############################################################################
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# Exporting to OpenMC settings.xml file
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###############################################################################
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# Instantiate a Settings object, set all runtime parameters, and export to XML
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settings_file = openmc.Settings()
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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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# Create an initial uniform spatial source distribution over fissionable zones
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uniform_dist = openmc.stats.Point(xyz=(0, 0, 0))
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settings_file.source = openmc.source.Source(space=uniform_dist)
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settings_file.export_to_xml()
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