mirror of
https://github.com/openmc-dev/openmc.git
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5111aa2621
Co-authored-by: Gavin Ridley <gavin.keith.ridley@gmail.com> Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
232 lines
9.4 KiB
Python
232 lines
9.4 KiB
Python
import os
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import numpy as np
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import openmc
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from tests.testing_harness import TolerantPyAPITestHarness
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class MGXSTestHarness(TolerantPyAPITestHarness):
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def _cleanup(self):
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super()._cleanup()
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f = 'mgxs.h5'
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if os.path.exists(f):
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os.remove(f)
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def random_ray_model() -> openmc.Model:
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###############################################################################
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# Create multigroup data
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# Instantiate the energy group data
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group_edges = [1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6]
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groups = openmc.mgxs.EnergyGroups(group_edges)
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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 = openmc.MGXSLibrary(groups)
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mg_cross_sections.add_xsdatas([uo2_xsdata, h2o_xsdata])
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mg_cross_sections.export_to_hdf5()
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###############################################################################
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# Create materials for the problem
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# Instantiate some Materials and register the appropriate macroscopic data
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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')
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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('LWTR')
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# Instantiate a Materials collection and export to XML
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materials = openmc.Materials([uo2, water])
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materials.cross_sections = "mgxs.h5"
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###############################################################################
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# Define problem geometry
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########################################
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# Define an unbounded pincell universe
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pitch = 1.26
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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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inner_ring_a = openmc.ZCylinder(r=0.33, name='inner ring a')
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inner_ring_b = openmc.ZCylinder(r=0.45, name='inner ring b')
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outer_ring_a = openmc.ZCylinder(r=0.60, name='outer ring a')
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outer_ring_b = openmc.ZCylinder(r=0.69, name='outer ring b')
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# Instantiate Cells
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fuel_a = openmc.Cell(fill=uo2, region=-inner_ring_a, name='fuel inner a')
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fuel_b = openmc.Cell(fill=uo2, region=+inner_ring_a & -inner_ring_b, name='fuel inner b')
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fuel_c = openmc.Cell(fill=uo2, region=+inner_ring_b & -fuel_or, name='fuel inner c')
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moderator_a = openmc.Cell(fill=water, region=+fuel_or & -outer_ring_a, name='moderator inner a')
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moderator_b = openmc.Cell(fill=water, region=+outer_ring_a & -outer_ring_b, name='moderator outer b')
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moderator_c = openmc.Cell(fill=water, region=+outer_ring_b, name='moderator outer c')
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# Create pincell universe
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pincell_base = openmc.Universe()
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# Register Cells with Universe
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pincell_base.add_cells([fuel_a, fuel_b, fuel_c, moderator_a, moderator_b, moderator_c])
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# Create planes for azimuthal sectors
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azimuthal_planes = []
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for i in range(8):
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angle = 2 * i * openmc.pi / 8
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normal_vector = (-openmc.sin(angle), openmc.cos(angle), 0)
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azimuthal_planes.append(openmc.Plane(a=normal_vector[0], b=normal_vector[1], c=normal_vector[2], d=0))
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# Create a cell for each azimuthal sector
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azimuthal_cells = []
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for i in range(8):
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azimuthal_cell = openmc.Cell(name=f'azimuthal_cell_{i}')
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azimuthal_cell.fill = pincell_base
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azimuthal_cell.region = +azimuthal_planes[i] & -azimuthal_planes[(i+1) % 8]
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azimuthal_cells.append(azimuthal_cell)
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# Create a geometry with the azimuthal universes
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pincell = openmc.Universe(cells=azimuthal_cells)
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########################################
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# Define a moderator lattice universe
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moderator_infinite = openmc.Cell(fill=water, name='moderator infinite')
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mu = openmc.Universe()
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mu.add_cells([moderator_infinite])
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lattice = openmc.RectLattice()
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lattice.lower_left = [-pitch/2.0, -pitch/2.0]
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lattice.pitch = [pitch/10.0, pitch/10.0]
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lattice.universes = np.full((10, 10), mu)
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mod_lattice_cell = openmc.Cell(fill=lattice)
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mod_lattice_uni = openmc.Universe()
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mod_lattice_uni.add_cells([mod_lattice_cell])
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########################################
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# Define 2x2 outer lattice
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lattice2x2 = openmc.RectLattice()
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lattice2x2.lower_left = (-pitch, -pitch)
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lattice2x2.pitch = (pitch, pitch)
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lattice2x2.universes = [
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[pincell, pincell],
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[pincell, mod_lattice_uni]
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]
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########################################
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# Define cell containing lattice and other stuff
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box = openmc.model.RectangularPrism(pitch*2, pitch*2, boundary_type='reflective')
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assembly = openmc.Cell(fill=lattice2x2, region=-box, name='assembly')
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# Create a geometry with the top-level cell
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geometry = openmc.Geometry([assembly])
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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 = openmc.Settings()
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settings.energy_mode = "multi-group"
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settings.batches = 10
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settings.inactive = 5
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settings.particles = 100
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# Create an initial uniform spatial source distribution over fissionable zones
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lower_left = (-pitch, -pitch, -1)
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upper_right = (pitch, pitch, 1)
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uniform_dist = openmc.stats.Box(lower_left, upper_right)
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rr_source = openmc.IndependentSource(space=uniform_dist)
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settings.random_ray['distance_active'] = 100.0
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settings.random_ray['distance_inactive'] = 20.0
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settings.random_ray['ray_source'] = rr_source
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###############################################################################
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# Define tallies
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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 = (2, 2)
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mesh.lower_left = (-pitch, -pitch)
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mesh.upper_right = (pitch, pitch)
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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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# Create an energy group filter as well
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energy_filter = openmc.EnergyFilter(group_edges)
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# Now use the mesh filter in a tally and indicate what scores are desired
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tally = openmc.Tally(name="Mesh tally")
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tally.filters = [mesh_filter, energy_filter]
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tally.scores = ['flux', 'fission', 'nu-fission']
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tally.estimator = 'analog'
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# Instantiate a Tallies collection and export to XML
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tallies = openmc.Tallies([tally])
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###############################################################################
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# Exporting to OpenMC model
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###############################################################################
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model = openmc.Model()
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model.geometry = geometry
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model.materials = materials
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model.settings = settings
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model.tallies = tallies
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return model
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def test_random_ray_basic():
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harness = MGXSTestHarness('statepoint.10.h5', random_ray_model())
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harness.main()
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