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Resolved conflict
This commit is contained in:
zoe prieto
commit
dc2a9a8129
221 changed files with 16404 additions and 15731 deletions
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@ -53,16 +53,19 @@ _MIN_E = 1.e-5
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_MAX_E = 20.e6
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def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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axis=None, sab_name=None, ce_cross_sections=None,
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mg_cross_sections=None, enrichment=None, plot_CE=True, orders=None,
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divisor_orders=None, **kwargs):
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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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**kwargs):
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"""Creates a figure of continuous-energy cross sections for this item.
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Parameters
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----------
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this : str or openmc.Material
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Object to source data from
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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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divisor_types : Iterable of values of PLOT_TYPES, optional
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@ -74,23 +77,18 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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temperature of 294K will be plotted. Note that the nearest
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temperature in the library for each nuclide will be used as opposed
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to using any interpolation.
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data_type : {'nuclide', 'element', 'material', 'macroscopic'}, optional
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Type of object to plot. If not specified, a guess is made based on the
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`this` argument.
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axis : matplotlib.axes, optional
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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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@ -101,7 +99,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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multi-group data.
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divisor_orders : Iterable of Integral, optional
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Same as orders, but for divisor_types
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**kwargs
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**kwargs :
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All keyword arguments are passed to
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:func:`matplotlib.pyplot.figure`.
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@ -117,27 +115,11 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=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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if data_type is None:
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if isinstance(this, openmc.Nuclide):
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data_type = 'nuclide'
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elif isinstance(this, openmc.Element):
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data_type = 'element'
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elif isinstance(this, openmc.Material):
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data_type = 'material'
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elif isinstance(this, openmc.Macroscopic):
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data_type = 'macroscopic'
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elif isinstance(this, str):
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if this[-1] in string.digits:
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data_type = 'nuclide'
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else:
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data_type = 'element'
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else:
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raise TypeError("Invalid type for plotting")
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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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E, data = calculate_cexs(this, data_type, types, temperature, sab_name,
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E, data = calculate_cexs(this, types, temperature, sab_name,
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ce_cross_sections, enrichment)
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if divisor_types:
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cv.check_length('divisor types', divisor_types, len(types))
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@ -160,13 +142,13 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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data = data_new
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else:
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# Calculate for MG cross sections
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E, data = calculate_mgxs(this, data_type, types, orders, temperature,
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E, data = calculate_mgxs(this, types, orders, temperature,
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mg_cross_sections, ce_cross_sections,
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enrichment)
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if divisor_types:
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cv.check_length('divisor types', divisor_types, len(types))
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Ediv, data_div = calculate_mgxs(this, data_type, divisor_types,
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Ediv, data_div = calculate_mgxs(this, divisor_types,
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divisor_orders, temperature,
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mg_cross_sections,
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ce_cross_sections, enrichment)
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@ -179,7 +161,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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# Generate the plot
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if axis is None:
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fig, ax = plt.subplots()
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fig, ax = plt.subplots(**kwargs)
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else:
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fig = None
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ax = axis
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@ -201,23 +183,30 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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ax.set_xlim(_MIN_E, _MAX_E)
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else:
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ax.set_xlim(E[-1], E[0])
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if divisor_types:
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if data_type == 'nuclide':
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ylabel = 'Nuclidic Microscopic Data'
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elif data_type == 'element':
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ylabel = 'Elemental Microscopic Data'
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elif data_type == 'material' or data_type == '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 data_type == 'nuclide':
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ylabel = 'Microscopic Cross Section [b]'
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elif data_type == 'element':
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ylabel = 'Elemental Cross Section [b]'
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elif data_type == 'material' or data_type == '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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ax.set_ylabel(ylabel)
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ax.legend(loc='best')
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name = this.name if data_type == 'material' else this
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name = this.name if isinstance(this, openmc.Material) else this
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if len(types) > 1:
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ax.set_title('Cross Sections for ' + name)
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else:
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@ -226,16 +215,16 @@ def plot_xs(this, types, divisor_types=None, temperature=294., data_type=None,
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return fig
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def calculate_cexs(this, data_type, types, temperature=294., sab_name=None,
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def calculate_cexs(this, types, temperature=294., sab_name=None,
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cross_sections=None, enrichment=None, ncrystal_cfg=None):
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"""Calculates continuous-energy cross sections of a requested type.
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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
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data_type : {'nuclide', 'element', 'material'}
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Type of object to plot
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this : str or 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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@ -262,40 +251,33 @@ def calculate_cexs(this, data_type, 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.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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if data_type == 'nuclide':
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if isinstance(this, str):
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nuc = openmc.Nuclide(this)
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if isinstance(this, str):
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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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nuc = this
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energy_grid, xs = _calculate_cexs_nuclide(nuc, types, temperature,
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sab_name, cross_sections,
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ncrystal_cfg)
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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 data_type == 'element':
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if isinstance(this, str):
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elem = openmc.Element(this)
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else:
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elem = this
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energy_grid, data = _calculate_cexs_elem_mat(elem, types, temperature,
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cross_sections, sab_name,
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enrichment)
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elif data_type == 'material':
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cv.check_type('this', this, openmc.Material)
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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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ncrystal_cfg
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)
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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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else:
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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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raise TypeError("Invalid type")
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return energy_grid, data
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@ -306,7 +288,7 @@ 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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@ -516,8 +498,8 @@ 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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@ -559,20 +541,18 @@ 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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# Add NCrystal cfg string if it exists
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ncrystal_cfg = this.ncrystal_cfg
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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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@ -600,8 +580,9 @@ def _calculate_cexs_elem_mat(this, types, temperature=294.,
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name = nuclide[0]
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nuc = nuclide[1]
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sab_tab = sabs[name]
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temp_E, temp_xs = calculate_cexs(nuc, 'nuclide', types, T, sab_tab,
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cross_sections, ncrystal_cfg=ncrystal_cfg)
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temp_E, temp_xs = calculate_cexs(nuc, types, T, sab_tab, cross_sections,
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ncrystal_cfg=ncrystal_cfg
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)
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E.append(temp_E)
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# Since the energy grids are different, store the cross sections as
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# a tabulated function so they can be calculated on any grid needed.
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@ -628,7 +609,7 @@ def _calculate_cexs_elem_mat(this, types, temperature=294.,
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return energy_grid, data
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def calculate_mgxs(this, data_type, types, orders=None, temperature=294.,
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def calculate_mgxs(this, types, orders=None, temperature=294.,
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cross_sections=None, ce_cross_sections=None,
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enrichment=None):
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"""Calculates multi-group cross sections of a requested type.
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@ -640,9 +621,7 @@ def calculate_mgxs(this, data_type, types, orders=None, temperature=294.,
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Parameters
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----------
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this : str or openmc.Material
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Object to source data from
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data_type : {'nuclide', 'element', 'material', 'macroscopic'}
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Type of object to plot
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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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orders : Iterable of Integral, optional
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@ -658,7 +637,6 @@ def calculate_mgxs(this, data_type, 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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@ -682,13 +660,13 @@ def calculate_mgxs(this, data_type, 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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if data_type in ('nuclide', 'macroscopic'):
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mgxs = _calculate_mgxs_nuc_macro(this, types, library, orders,
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temperature)
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elif data_type in ('element', '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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@ -720,7 +698,7 @@ 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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@ -858,8 +836,8 @@ 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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@ -908,7 +886,7 @@ 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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|
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