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removed most deprecated types
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shimwell
parent
1a9449debc
commit
17e9918f5d
2 changed files with 55 additions and 76 deletions
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@ -53,6 +53,9 @@ _MIN_E = 1.e-5
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_MAX_E = 20.e6
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ELEMENT_NAMES = list(openmc.data.ELEMENT_SYMBOL.values())[1:]
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def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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sab_name=None, ce_cross_sections=None, mg_cross_sections=None,
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enrichment=None, plot_CE=True, orders=None, divisor_orders=None):
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@ -60,7 +63,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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Parameters
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----------
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this : {str, openmc.Nuclide, openmc.Element, openmc.Macroscopic, openmc.Material}
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this : {str, openmc.Material}
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Object to source data from. Nuclides and Elements can be input as a str
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types : Iterable of values of PLOT_TYPES
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The type of cross sections to include in the plot.
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@ -77,16 +80,14 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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A previously generated axis to use for plotting. If not specified,
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a new axis and figure will be generated.
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sab_name : str, optional
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Name of S(a,b) library to apply to MT=2 data when applicable; only used
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for items which are instances of openmc.Element or openmc.Nuclide
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Name of S(a,b) library to apply to MT=2 data when applicable.
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ce_cross_sections : str, optional
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Location of cross_sections.xml file. Default is None.
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mg_cross_sections : str, optional
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Location of MGXS HDF5 Library file. Default is None.
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enrichment : float, optional
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Enrichment for U235 in weight percent. For example, input 4.95 for
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4.95 weight percent enriched U. Default is None. This is only used for
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items which are instances of openmc.Element
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4.95 weight percent enriched U. Default is None.
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plot_CE : bool, optional
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Denotes whether or not continuous-energy will be plotted. Defaults to
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plotting the continuous-energy data.
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@ -110,7 +111,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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import matplotlib.pyplot as plt
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cv.check_type("plot_CE", plot_CE, bool)
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cv.check_type("this", this, (str, openmc.Nuclide, openmc.Element, openmc.Material))
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cv.check_type("this", this, (str, openmc.Material))
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if plot_CE:
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# Calculate for the CE cross sections
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@ -179,28 +180,23 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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else:
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ax.set_xlim(E[-1], E[0])
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if isinstance(this, str):
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# first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry
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if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]:
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this = openmc.Element(this)
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else:
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this = openmc.Nuclide(this)
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if divisor_types:
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if isinstance(this, openmc.Nuclide):
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ylabel = 'Nuclidic Microscopic Data'
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elif isinstance(this, openmc.Element):
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ylabel = 'Elemental Microscopic Data'
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elif isinstance(this, openmc.Material) or isinstance(this, openmc.Macroscopic):
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if isinstance(this, str):
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if this in ELEMENT_NAMES:
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ylabel = 'Elemental Microscopic Data'
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else:
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ylabel = 'Nuclide Microscopic Data'
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elif isinstance(this, openmc.Material):
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ylabel = 'Macroscopic Data'
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else:
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raise TypeError("Invalid type for plotting")
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else:
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if isinstance(this, openmc.Nuclide):
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ylabel = 'Microscopic Cross Section [b]'
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elif isinstance(this, openmc.Element):
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ylabel = 'Elemental Cross Section [b]'
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elif isinstance(this, openmc.Material) or isinstance(this, openmc.Macroscopic):
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if isinstance(this, str):
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if this in ELEMENT_NAMES:
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ylabel = 'Elemental Cross Section [b]'
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else:
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ylabel = 'Microscopic Cross Section [b]'
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elif isinstance(this, openmc.Material):
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ylabel = 'Macroscopic Cross Section [1/cm]'
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else:
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raise TypeError("Invalid type for plotting")
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@ -221,8 +217,9 @@ def calculate_cexs(this, types, temperature=294., sab_name=None,
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Parameters
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----------
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this : {str, openmc.Nuclide, openmc.Element, openmc.Material}
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Object to source data from. Nuclides and Elements can be input as a str
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this : {str, openmc.Material}
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Object to source data from. Nuclides and Elements should be input as a
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str
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types : Iterable of values of PLOT_TYPES
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The type of cross sections to calculate
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temperature : float, optional
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@ -249,44 +246,37 @@ def calculate_cexs(this, types, temperature=294., sab_name=None,
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"""
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# Check types
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cv.check_type('this', this, (str, openmc.Nuclide, openmc.Element, openmc.Material))
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cv.check_type('this', this, (str, openmc.Material))
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cv.check_type('temperature', temperature, Real)
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if sab_name:
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cv.check_type('sab_name', sab_name, str)
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if enrichment:
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cv.check_type('enrichment', enrichment, Real)
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# this is a nuclide or element if it is a string
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if isinstance(this, str):
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# first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry
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if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]:
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this = openmc.Element(this)
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if this in ELEMENT_NAMES:
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energy_grid, data = _calculate_cexs_elem_mat(
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this, types, temperature, cross_sections, sab_name, enrichment
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)
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else:
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this = openmc.Nuclide(this)
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energy_grid, xs = _calculate_cexs_nuclide(
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this, types, temperature, sab_name, cross_sections
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)
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if isinstance(this, openmc.Nuclide):
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energy_grid, xs = _calculate_cexs_nuclide(this, types, temperature,
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sab_name, cross_sections)
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# Convert xs (Iterable of Callable) to a grid of cross section values
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# calculated on the points in energy_grid for consistency with the
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# element and material functions.
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data = np.zeros((len(types), len(energy_grid)))
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for line in range(len(types)):
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data[line, :] = xs[line](energy_grid)
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elif isinstance(this, openmc.Element):
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energy_grid, data = _calculate_cexs_elem_mat(this, types, temperature,
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cross_sections, sab_name,
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enrichment)
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# Convert xs (Iterable of Callable) to a grid of cross section values
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# calculated on the points in energy_grid for consistency with the
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# element and material functions.
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data = np.zeros((len(types), len(energy_grid)))
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for line in range(len(types)):
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data[line, :] = xs[line](energy_grid)
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elif isinstance(this, openmc.Material):
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energy_grid, data = _calculate_cexs_elem_mat(this, types, temperature,
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cross_sections)
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else:
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msg = (
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f"{this} is an invalid type, acceptable types are str, "
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"openmc.Nuclide, openmc.Element, openmc.Material."
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f"{this} is an invalid type, acceptable types are str, openmc.Material."
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)
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raise TypeError(msg)
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@ -299,7 +289,7 @@ def _calculate_cexs_nuclide(this, types, temperature=294., sab_name=None,
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Parameters
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----------
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this : openmc.Nuclide
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this : str
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Nuclide object to source data from
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types : Iterable of str or Integral
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The type of cross sections to calculate; values can either be those
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@ -496,8 +486,8 @@ def _calculate_cexs_elem_mat(this, types, temperature=294.,
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Parameters
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----------
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this : openmc.Material or openmc.Element
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Object to source data from
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this : openmc.Material or str
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Object to source data from. Element can be input as str
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types : Iterable of values of PLOT_TYPES
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The type of cross sections to calculate
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temperature : float, optional
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@ -538,18 +528,16 @@ def _calculate_cexs_elem_mat(this, types, temperature=294.,
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# Expand elements in to nuclides with atomic densities
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nuc_fractions = this.get_nuclide_atom_densities()
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# Create a dict of [nuclide name] = nuclide object to carry forward
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# with a common nuclides format between openmc.Material and
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# openmc.Element objects
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# with a common nuclides format between openmc.Material and Elements
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nuclides = {nuclide: nuclide for nuclide in nuc_fractions}
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else:
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# Expand elements in to nuclides with atomic densities
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nuclides = this.expand(1., 'ao', enrichment=enrichment,
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nuclides = openmc.Element(this).expand(1., 'ao', enrichment=enrichment,
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cross_sections=cross_sections)
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# For ease of processing split out the nuclide and its fraction
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nuc_fractions = {nuclide[0]: nuclide[1] for nuclide in nuclides}
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# Create a dict of [nuclide name] = nuclide object to carry forward
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# with a common nuclides format between openmc.Material and
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# openmc.Element objects
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# with a common nuclides format between openmc.Material and Elements
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nuclides = {nuclide[0]: nuclide[0] for nuclide in nuclides}
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# Identify the nuclides which have S(a,b) data
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@ -615,7 +603,7 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
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Parameters
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----------
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this : {str, openmc.Nuclide, openmc.Element, openmc.Macroscopic, openmc.Material}
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this : {str, openmc.Material}
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Object to source data from. Nuclides and Elements can be input as a str
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types : Iterable of values of PLOT_TYPES_MGXS
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The type of cross sections to calculate
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@ -632,7 +620,6 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
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Location of MGXS HDF5 Library file. Default is None.
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ce_cross_sections : str, optional
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Location of continuous-energy cross_sections.xml file. Default is None.
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This is used only for expanding an openmc.Element object passed as this
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enrichment : float, optional
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Enrichment for U235 in weight percent. For example, input 4.95 for
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4.95 weight percent enriched U. Default is None
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@ -656,21 +643,13 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
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cv.check_type("cross_sections", cross_sections, str)
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library = openmc.MGXSLibrary.from_hdf5(cross_sections)
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# this is a nuclide or element if it is a string
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if isinstance(this, str):
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# first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry
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if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]:
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this = openmc.Element(this)
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else:
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this = openmc.Nuclide(this)
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if isinstance(this, openmc.Nuclide) or isinstance(this, openmc.Macroscopic):
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mgxs = _calculate_mgxs_nuc_macro(this, types, library, orders,
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temperature)
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elif isinstance(this, openmc.Element) or isinstance(this, openmc.Material):
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if this in ELEMENT_NAMES or isinstance(this, openmc.Material):
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mgxs = _calculate_mgxs_elem_mat(this, types, library, orders,
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temperature, ce_cross_sections,
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enrichment)
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elif isinstance(this, str):
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mgxs = _calculate_mgxs_nuc_macro(this, types, library, orders,
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temperature)
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else:
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raise TypeError("Invalid type")
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@ -702,7 +681,7 @@ def _calculate_mgxs_nuc_macro(this, types, library, orders=None,
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Parameters
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----------
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this : openmc.Nuclide or openmc.Macroscopic
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this : str
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Object to source data from
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types : Iterable of str
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The type of cross sections to calculate; values can either be those
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@ -840,8 +819,8 @@ def _calculate_mgxs_elem_mat(this, types, library, orders=None,
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Parameters
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----------
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this : openmc.Element or openmc.Material
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Object to source data from
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this : str or openmc.Material
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Object to source data from. Elements can be input as a str
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types : Iterable of str
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The type of cross sections to calculate; values can either be those
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in openmc.PLOT_TYPES_MGXS
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@ -890,7 +869,7 @@ def _calculate_mgxs_elem_mat(this, types, library, orders=None,
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else:
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T = temperature
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# Expand elements in to nuclides with atomic densities
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nuclides = this.expand(100., 'ao', enrichment=enrichment,
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nuclides = openmc.Element(this).expand(100., 'ao', enrichment=enrichment,
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cross_sections=ce_cross_sections)
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# For ease of processing split out nuc and nuc_fractions
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@ -32,7 +32,7 @@ def test_calculate_cexs_elem_mat_sab(test_mat):
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assert len(data[0]) == len(energy_grid)
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@pytest.mark.parametrize("this", ["Li", "Li6", openmc.Nuclide('Li6'), openmc.Element('Li')])
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@pytest.mark.parametrize("this", ["Li", "Li6"])
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def test_calculate_cexs_with_nuclide_and_element(this):
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# single type (reaction)
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energy_grid, data = openmc.plotter.calculate_cexs(
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@ -72,7 +72,7 @@ def test_calculate_cexs_with_materials(test_mat):
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assert len(data[0]) == len(energy_grid)
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@pytest.mark.parametrize("this", ["Be", "Be9", openmc.Nuclide('Be9'), openmc.Element('Be')])
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@pytest.mark.parametrize("this", ["Be", "Be9"])
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def test_plot_xs(this):
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assert isinstance(openmc.plotter.plot_xs(this, types=['total']), Figure)
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