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176 lines
7.1 KiB
Python
176 lines
7.1 KiB
Python
import numpy as np
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import openmc
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import openmc.mgxs
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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 mgxs.h5 file
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###############################################################################
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# Instantiate the energy group data
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groups = openmc.mgxs.EnergyGroups(group_edges=[
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1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6])
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# Instantiate the 7-group (C5G7) cross section data
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uo2_xsdata = openmc.XSdata('UO2', groups)
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uo2_xsdata.order = 0
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uo2_xsdata.set_total(
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[0.1779492, 0.3298048, 0.4803882, 0.5543674, 0.3118013, 0.3951678,
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0.5644058])
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uo2_xsdata.set_absorption([8.0248E-03, 3.7174E-03, 2.6769E-02, 9.6236E-02,
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3.0020E-02, 1.1126E-01, 2.8278E-01])
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scatter_matrix = np.array(
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[[[0.1275370, 0.0423780, 0.0000094, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.3244560, 0.0016314, 0.0000000, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.4509400, 0.0026792, 0.0000000, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.4525650, 0.0055664, 0.0000000, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.0001253, 0.2714010, 0.0102550, 0.0000000],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0012968, 0.2658020, 0.0168090],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0085458, 0.2730800]]])
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scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
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uo2_xsdata.set_scatter_matrix(scatter_matrix)
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uo2_xsdata.set_fission([7.21206E-03, 8.19301E-04, 6.45320E-03,
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1.85648E-02, 1.78084E-02, 8.30348E-02,
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2.16004E-01])
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uo2_xsdata.set_nu_fission([2.005998E-02, 2.027303E-03, 1.570599E-02,
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4.518301E-02, 4.334208E-02, 2.020901E-01,
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5.257105E-01])
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uo2_xsdata.set_chi([5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07, 0.0000E+00,
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0.0000E+00, 0.0000E+00])
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h2o_xsdata = openmc.XSdata('LWTR', groups)
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h2o_xsdata.order = 0
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h2o_xsdata.set_total([0.15920605, 0.412969593, 0.59030986, 0.58435,
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0.718, 1.2544497, 2.650379])
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h2o_xsdata.set_absorption([6.0105E-04, 1.5793E-05, 3.3716E-04,
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1.9406E-03, 5.7416E-03, 1.5001E-02,
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3.7239E-02])
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scatter_matrix = np.array(
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[[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0.0000000],
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[0.0000000, 0.2823340, 0.1299400, 0.0006234, 0.0000480, 0.0000074, 0.0000010],
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[0.0000000, 0.0000000, 0.3452560, 0.2245700, 0.0169990, 0.0026443, 0.0005034],
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[0.0000000, 0.0000000, 0.0000000, 0.0910284, 0.4155100, 0.0637320, 0.0121390],
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[0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200],
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[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]])
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scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
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h2o_xsdata.set_scatter_matrix(scatter_matrix)
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mg_cross_sections_file = openmc.MGXSLibrary(groups)
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mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata])
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mg_cross_sections_file.export_to_hdf5()
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###############################################################################
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Instantiate some Macroscopic Data
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uo2_data = openmc.Macroscopic('UO2')
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h2o_data = openmc.Macroscopic('LWTR')
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# Instantiate some Materials and register the appropriate Macroscopic objects
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uo2 = openmc.Material(material_id=1, name='UO2 fuel')
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uo2.set_density('macro', 1.0)
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uo2.add_macroscopic(uo2_data)
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water = openmc.Material(material_id=2, name='Water')
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water.set_density('macro', 1.0)
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water.add_macroscopic(h2o_data)
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# Instantiate a Materials collection and export to XML
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materials_file = openmc.Materials([uo2, water])
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materials_file.cross_sections = "./mgxs.h5"
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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=0.54, name='Fuel OR')
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left = openmc.XPlane(surface_id=4, x0=-0.63, name='left')
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right = openmc.XPlane(surface_id=5, x0=0.63, name='right')
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bottom = openmc.YPlane(surface_id=6, y0=-0.63, name='bottom')
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top = openmc.YPlane(surface_id=7, y0=0.63, 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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moderator = openmc.Cell(cell_id=2, name='cell 2')
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# Use surface half-spaces to define regions
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fuel.region = -fuel_or
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moderator.region = +fuel_or & +left & -right & +bottom & -top
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# Register Materials with Cells
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fuel.fill = uo2
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moderator.fill = 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, moderator])
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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.energy_mode = "multi-group"
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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 = [-0.63, -0.63, -1, 0.63, 0.63, 1]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
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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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###############################################################################
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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.Mesh(mesh_id=1)
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mesh.type = 'regular'
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mesh.dimension = [100, 100, 1]
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mesh.lower_left = [-0.63, -0.63, -1.e50]
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mesh.upper_right = [0.63, 0.63, 1.e50]
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# Instantiate some tally Filters
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energy_filter = openmc.EnergyFilter([1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6,
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1.0e6, 20.0e6])
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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, register all Tallies, 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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