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Update multigroup example
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Paul Romano
parent
03a349916d
commit
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1 changed files with 53 additions and 78 deletions
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@ -1,20 +1,12 @@
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from math import log10
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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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# Create multigroup data
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# Instantiate the energy group data
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groups = openmc.mgxs.EnergyGroups(group_edges=[
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@ -69,107 +61,90 @@ 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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# Create materials for the problem
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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 = openmc.Material(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 = openmc.Material(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.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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# Define problem geometry
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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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# Create a surface for the fuel outer radius
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fuel_or = openmc.ZCylinder(r=0.54, name='Fuel OR')
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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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# Create a region represented as the inside of a rectangular prism
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pitch = 1.26
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box = openmc.rectangular_prism(pitch, pitch, 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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fuel = openmc.Cell(fill=uo2, region=-fuel_or, name='fuel')
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moderator = openmc.Cell(fill=water, region=+fuel_or & box, name='moderator')
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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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# Create a geometry with the two cells and export to XML
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geometry = openmc.Geometry([fuel, moderator])
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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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# Define problem settings
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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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settings = openmc.Settings()
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settings.energy_mode = "multi-group"
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settings.batches = 100
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settings.inactive = 10
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settings.particles = 1000
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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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lower_left = (-pitch/2, -pitch/2, -1)
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upper_right = (pitch/2, pitch/2, 1)
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uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
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settings.source = openmc.source.Source(space=uniform_dist)
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settings.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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# Define tallies
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# Instantiate a tally mesh
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mesh = openmc.RegularMesh(mesh_id=1)
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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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# Create a mesh that will be used for tallying
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mesh = openmc.RegularMesh()
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mesh.dimension = (100, 100)
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mesh.lower_left = (-pitch/2, -pitch/2)
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mesh.upper_right = (pitch/2, pitch/2)
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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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# Create a mesh filter that can be used in a tally
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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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# Now use the mesh filter in a tally and indicate what scores are desired
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mesh_tally = openmc.Tally(name="Mesh tally")
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mesh_tally.filters = [mesh_filter]
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mesh_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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# Let's also create a tally to get the flux energy spectrum. We start by
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# creating an energy filter
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e_min, e_max = 1e-5, 20.0e6
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groups = 500
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energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
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energy_filter = openmc.EnergyFilter(energies)
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spectrum_tally = openmc.Tally(name="Flux spectrum")
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spectrum_tally.filters = [energy_filter]
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spectrum_tally.scores = ['flux']
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# Instantiate a Tallies collection and export to XML
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tallies = openmc.Tallies([mesh_tally, spectrum_tally])
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tallies.export_to_xml()
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