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138 lines
5 KiB
Python
138 lines
5 KiB
Python
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 = 100
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inactive = 10
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particles = 1000
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###############################################################################
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Materials and register the appropriate Nuclides
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uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
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uo2.set_density('g/cm3', 10.29769)
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uo2.add_element('U', 1., enrichment=2.4)
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uo2.add_element('O', 2.)
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helium = openmc.Material(material_id=2, name='Helium for gap')
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helium.set_density('g/cm3', 0.001598)
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helium.add_element('He', 2.4044e-4)
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zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
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zircaloy.set_density('g/cm3', 6.55)
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zircaloy.add_element('Sn', 0.014 , 'wo')
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zircaloy.add_element('Fe', 0.00165, 'wo')
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zircaloy.add_element('Cr', 0.001 , 'wo')
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zircaloy.add_element('Zr', 0.98335, 'wo')
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borated_water = openmc.Material(material_id=4, name='Borated water')
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borated_water.set_density('g/cm3', 0.740582)
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borated_water.add_element('B', 4.0e-5)
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borated_water.add_element('H', 5.0e-2)
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borated_water.add_element('O', 2.4e-2)
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borated_water.add_s_alpha_beta('c_H_in_H2O')
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# Instantiate a Materials collection and export to XML
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materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
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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 ZCylinder surfaces
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fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='Fuel OR')
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clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.40005, name='Clad IR')
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clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=0.45720, name='Clad OR')
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left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
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right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
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bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
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top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
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left.boundary_type = 'reflective'
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right.boundary_type = 'reflective'
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top.boundary_type = 'reflective'
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bottom.boundary_type = 'reflective'
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# Instantiate Cells
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fuel = openmc.Cell(cell_id=1, name='cell 1')
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gap = openmc.Cell(cell_id=2, name='cell 2')
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clad = openmc.Cell(cell_id=3, name='cell 3')
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water = openmc.Cell(cell_id=4, name='cell 4')
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# Use surface half-spaces to define regions
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fuel.region = -fuel_or
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gap.region = +fuel_or & -clad_ir
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clad.region = +clad_ir & -clad_or
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water.region = +clad_or & +left & -right & +bottom & -top
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# Register Materials with Cells
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fuel.fill = uo2
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gap.fill = helium
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clad.fill = zircaloy
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water.fill = borated_water
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# Instantiate Universe
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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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root.add_cells([fuel, gap, clad, water])
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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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bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
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settings_file.source = openmc.source.Source(space=uniform_dist)
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entropy_mesh = openmc.RegularMesh()
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entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
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entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
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entropy_mesh.dimension = [10, 10, 1]
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settings_file.entropy_mesh = entropy_mesh
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settings_file.export_to_xml()
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###############################################################################
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# Exporting to OpenMC tallies.xml file
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###############################################################################
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# Instantiate a tally mesh
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mesh = openmc.RegularMesh()
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mesh.dimension = [100, 100, 1]
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mesh.lower_left = [-0.62992, -0.62992, -1.e50]
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mesh.upper_right = [0.62992, 0.62992, 1.e50]
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# Instantiate some tally Filters
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energy_filter = openmc.EnergyFilter([0., 4., 20.e6])
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mesh_filter = openmc.MeshFilter(mesh)
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# Instantiate the Tally
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tally = openmc.Tally(tally_id=1, name='tally 1')
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tally.filters = [energy_filter, mesh_filter]
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tally.scores = ['flux', 'fission', 'nu-fission']
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# Instantiate a Tallies collection and export to XML
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tallies_file = openmc.Tallies([tally])
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tallies_file.export_to_xml()
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