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Replacing old slab_mg example model with new one that uses a library that follows the library standard. Also had to update tests accordingly.
This commit is contained in:
Adam G Nelson
parent
c7fa8c05ce
commit
c1652bbdec
25 changed files with 359 additions and 675 deletions
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@ -1,3 +1,5 @@
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from numbers import Integral
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import numpy as np
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import openmc
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@ -538,20 +540,20 @@ def pwr_assembly():
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return model
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def slab_mg(reps=None, as_macro=True):
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"""Create a one-group, 1D slab model.
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def slab_mg(num_regions=1, mat_names=None, mgxslib_name='2g.h5'):
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"""Create a 1D slab model.
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Parameters
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----------
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reps : list, optional
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List of angular representations. Each item corresponds to materials and
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dictates the angular representation of the multi-group cross
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sections---isotropic ('iso') or angle-dependent ('ang'), and if Legendre
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scattering or tabular scattering ('mu') is used. Thus, items can be
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'ang', 'ang_mu', 'iso', or 'iso_mu'.
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num_regions : int, optional
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Number of regions in the problem, each with a unique MGXS dataset.
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Defaults to 1.
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as_macro : bool, optional
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Whether :class:`openmc.Macroscopic` is used
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mat_names : Iterable of str, optional
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List of the material names to use; defaults to ['mat_1', 'mat_2',...].
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mgxslib_name : str, optional
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MGXS Library file to use; defaults to '2g.h5'.
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Returns
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-------
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@ -559,71 +561,82 @@ def slab_mg(reps=None, as_macro=True):
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One-group, 1D slab model
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"""
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openmc.check_type('num_regions', num_regions, Integral)
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openmc.check_greater_than('num_regions', num_regions, 0)
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if mat_names is not None:
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openmc.check_length('mat_names', mat_names, num_regions)
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openmc.check_iterable_type('mat_names', mat_names, str)
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else:
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mat_names = []
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for i in range(num_regions):
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mat_names.append('mat_' + str(i + 1))
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# # Make Materials
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materials_file = openmc.Materials()
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macros = []
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mats = []
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for i in range(len(mat_names)):
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macros.append(openmc.Macroscopic('mat_' + str(i + 1)))
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mats.append(openmc.Material(name=mat_names[i]))
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mats[-1].set_density('macro', 1.0)
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mats[-1].add_macroscopic(macros[-1])
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materials_file += mats
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materials_file.cross_sections = mgxslib_name
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# # Make Geometry
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rad_outer = 929.45
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# Set a cell boundary to exist for every material above (exclude the 0)
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rads = np.linspace(0., rad_outer, len(mats) + 1, endpoint=True)[1:]
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# Instantiate Universe
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root = openmc.Universe(universe_id=0, name='root universe')
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cells = []
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surfs = []
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surfs.append(openmc.XPlane(x0=0., boundary_type='reflective'))
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for r, rad in enumerate(rads):
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if r == len(rads) - 1:
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surfs.append(openmc.XPlane(x0=rad, boundary_type='vacuum'))
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else:
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surfs.append(openmc.XPlane(x0=rad))
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# Instantiate Cells
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cells = []
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for c in range(len(surfs) - 1):
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cells.append(openmc.Cell())
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cells[-1].region = (+surfs[c] & -surfs[c + 1])
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cells[-1].fill = mats[c]
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# Register Cells with Universe
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root.add_cells(cells)
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# Instantiate a Geometry, register the root Universe, and export to XML
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geometry_file = openmc.Geometry(root)
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# # Make Settings
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# Instantiate a Settings object, set all runtime parameters
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settings_file = openmc.Settings()
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settings_file.energy_mode = "multi-group"
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settings_file.tabular_legendre = {'enable': False}
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settings_file.batches = 10
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settings_file.inactive = 5
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settings_file.particles = 1000
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# Build source distribution
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INF = 1000.
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bounds = [0., -INF, -INF, rads[0], INF, INF]
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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.output = {'summary': False}
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model = openmc.model.Model()
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# Define materials needed for 1D/1G slab problem
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mat_names = ['uo2', 'clad', 'lwtr']
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mgxs_reps = ['ang', 'ang_mu', 'iso', 'iso_mu']
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if reps is None:
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reps = mgxs_reps
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xs = []
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i = 0
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for mat in mat_names:
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for rep in reps:
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i += 1
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name = mat + '_' + rep
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xs.append(name)
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if as_macro:
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m = openmc.Material(name=str(i))
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m.set_density('macro', 1.)
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m.add_macroscopic(name)
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else:
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m = openmc.Material(name=str(i))
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m.set_density('atom/b-cm', 1.)
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m.add_nuclide(name, 1.0, 'ao')
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model.materials.append(m)
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# Define the materials file
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model.xs_data = xs
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model.materials.cross_sections = "../../1d_mgxs.h5"
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# Define surfaces.
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# Assembly/Problem Boundary
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left = openmc.XPlane(x0=0.0, boundary_type='reflective')
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right = openmc.XPlane(x0=10.0, boundary_type='reflective')
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bottom = openmc.YPlane(y0=0.0, boundary_type='reflective')
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top = openmc.YPlane(y0=10.0, boundary_type='reflective')
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# for each material add a plane
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planes = [openmc.ZPlane(z0=0.0, boundary_type='reflective')]
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dz = round(5. / float(len(model.materials)), 4)
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for i in range(len(model.materials) - 1):
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planes.append(openmc.ZPlane(z0=dz * float(i + 1)))
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planes.append(openmc.ZPlane(z0=5.0, boundary_type='reflective'))
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# Define cells for each material
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model.geometry.root_universe = openmc.Universe(name='root universe')
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xy = +left & -right & +bottom & -top
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for i, mat in enumerate(model.materials):
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c = openmc.Cell(fill=mat, region=xy & +planes[i] & -planes[i + 1])
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model.geometry.root_universe.add_cell(c)
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model.settings.batches = 10
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model.settings.inactive = 5
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model.settings.particles = 100
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model.settings.source = openmc.Source(space=openmc.stats.Box(
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[0.0, 0.0, 0.0], [10.0, 10.0, 5.]))
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model.settings.energy_mode = "multi-group"
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plot = openmc.Plot()
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plot.filename = 'mat'
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plot.origin = (5.0, 5.0, 2.5)
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plot.width = (2.5, 2.5)
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plot.basis = 'xz'
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plot.pixels = (3000, 3000)
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plot.color_by = 'material'
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model.plots.append(plot)
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model.geometry = geometry_file
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model.materials = materials_file
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model.settings = settings_file
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model.xs_data = macros
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return model
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