From 17e9918f5d33188f5882559b073fee72dce3e195 Mon Sep 17 00:00:00 2001 From: shimwell Date: Tue, 21 Mar 2023 21:51:40 +0000 Subject: [PATCH] removed most deprecated types --- openmc/plotter.py | 127 +++++++++++++------------------ tests/unit_tests/test_plotter.py | 4 +- 2 files changed, 55 insertions(+), 76 deletions(-) diff --git a/openmc/plotter.py b/openmc/plotter.py index 541b05f0b2..2bdcfe3519 100644 --- a/openmc/plotter.py +++ b/openmc/plotter.py @@ -53,6 +53,9 @@ _MIN_E = 1.e-5 _MAX_E = 20.e6 +ELEMENT_NAMES = list(openmc.data.ELEMENT_SYMBOL.values())[1:] + + def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, sab_name=None, ce_cross_sections=None, mg_cross_sections=None, enrichment=None, plot_CE=True, orders=None, divisor_orders=None): @@ -60,7 +63,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, Parameters ---------- - this : {str, openmc.Nuclide, openmc.Element, openmc.Macroscopic, openmc.Material} + this : {str, openmc.Material} Object to source data from. Nuclides and Elements can be input as a str types : Iterable of values of PLOT_TYPES The type of cross sections to include in the plot. @@ -77,16 +80,14 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, A previously generated axis to use for plotting. If not specified, a new axis and figure will be generated. sab_name : str, optional - Name of S(a,b) library to apply to MT=2 data when applicable; only used - for items which are instances of openmc.Element or openmc.Nuclide + Name of S(a,b) library to apply to MT=2 data when applicable. ce_cross_sections : str, optional Location of cross_sections.xml file. Default is None. mg_cross_sections : str, optional Location of MGXS HDF5 Library file. Default is None. enrichment : float, optional Enrichment for U235 in weight percent. For example, input 4.95 for - 4.95 weight percent enriched U. Default is None. This is only used for - items which are instances of openmc.Element + 4.95 weight percent enriched U. Default is None. plot_CE : bool, optional Denotes whether or not continuous-energy will be plotted. Defaults to plotting the continuous-energy data. @@ -110,7 +111,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, import matplotlib.pyplot as plt cv.check_type("plot_CE", plot_CE, bool) - cv.check_type("this", this, (str, openmc.Nuclide, openmc.Element, openmc.Material)) + cv.check_type("this", this, (str, openmc.Material)) if plot_CE: # Calculate for the CE cross sections @@ -179,28 +180,23 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, else: ax.set_xlim(E[-1], E[0]) - if isinstance(this, str): - # first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry - if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]: - this = openmc.Element(this) - else: - this = openmc.Nuclide(this) - if divisor_types: - if isinstance(this, openmc.Nuclide): - ylabel = 'Nuclidic Microscopic Data' - elif isinstance(this, openmc.Element): - ylabel = 'Elemental Microscopic Data' - elif isinstance(this, openmc.Material) or isinstance(this, openmc.Macroscopic): + if isinstance(this, str): + if this in ELEMENT_NAMES: + ylabel = 'Elemental Microscopic Data' + else: + ylabel = 'Nuclide Microscopic Data' + elif isinstance(this, openmc.Material): ylabel = 'Macroscopic Data' else: raise TypeError("Invalid type for plotting") else: - if isinstance(this, openmc.Nuclide): - ylabel = 'Microscopic Cross Section [b]' - elif isinstance(this, openmc.Element): - ylabel = 'Elemental Cross Section [b]' - elif isinstance(this, openmc.Material) or isinstance(this, openmc.Macroscopic): + if isinstance(this, str): + if this in ELEMENT_NAMES: + ylabel = 'Elemental Cross Section [b]' + else: + ylabel = 'Microscopic Cross Section [b]' + elif isinstance(this, openmc.Material): ylabel = 'Macroscopic Cross Section [1/cm]' else: raise TypeError("Invalid type for plotting") @@ -221,8 +217,9 @@ def calculate_cexs(this, types, temperature=294., sab_name=None, Parameters ---------- - this : {str, openmc.Nuclide, openmc.Element, openmc.Material} - Object to source data from. Nuclides and Elements can be input as a str + this : {str, openmc.Material} + Object to source data from. Nuclides and Elements should be input as a + str types : Iterable of values of PLOT_TYPES The type of cross sections to calculate temperature : float, optional @@ -249,44 +246,37 @@ def calculate_cexs(this, types, temperature=294., sab_name=None, """ # Check types - cv.check_type('this', this, (str, openmc.Nuclide, openmc.Element, openmc.Material)) + cv.check_type('this', this, (str, openmc.Material)) cv.check_type('temperature', temperature, Real) if sab_name: cv.check_type('sab_name', sab_name, str) if enrichment: cv.check_type('enrichment', enrichment, Real) - # this is a nuclide or element if it is a string if isinstance(this, str): - # first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry - if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]: - this = openmc.Element(this) + if this in ELEMENT_NAMES: + energy_grid, data = _calculate_cexs_elem_mat( + this, types, temperature, cross_sections, sab_name, enrichment + ) + else: - this = openmc.Nuclide(this) + energy_grid, xs = _calculate_cexs_nuclide( + this, types, temperature, sab_name, cross_sections + ) - if isinstance(this, openmc.Nuclide): - energy_grid, xs = _calculate_cexs_nuclide(this, types, temperature, - sab_name, cross_sections) - - # Convert xs (Iterable of Callable) to a grid of cross section values - # calculated on the points in energy_grid for consistency with the - # element and material functions. - data = np.zeros((len(types), len(energy_grid))) - for line in range(len(types)): - data[line, :] = xs[line](energy_grid) - - elif isinstance(this, openmc.Element): - energy_grid, data = _calculate_cexs_elem_mat(this, types, temperature, - cross_sections, sab_name, - enrichment) + # Convert xs (Iterable of Callable) to a grid of cross section values + # calculated on the points in energy_grid for consistency with the + # element and material functions. + data = np.zeros((len(types), len(energy_grid))) + for line in range(len(types)): + data[line, :] = xs[line](energy_grid) elif isinstance(this, openmc.Material): energy_grid, data = _calculate_cexs_elem_mat(this, types, temperature, cross_sections) else: msg = ( - f"{this} is an invalid type, acceptable types are str, " - "openmc.Nuclide, openmc.Element, openmc.Material." + f"{this} is an invalid type, acceptable types are str, openmc.Material." ) raise TypeError(msg) @@ -299,7 +289,7 @@ def _calculate_cexs_nuclide(this, types, temperature=294., sab_name=None, Parameters ---------- - this : openmc.Nuclide + this : str Nuclide object to source data from types : Iterable of str or Integral The type of cross sections to calculate; values can either be those @@ -496,8 +486,8 @@ def _calculate_cexs_elem_mat(this, types, temperature=294., Parameters ---------- - this : openmc.Material or openmc.Element - Object to source data from + this : openmc.Material or str + Object to source data from. Element can be input as str types : Iterable of values of PLOT_TYPES The type of cross sections to calculate temperature : float, optional @@ -538,18 +528,16 @@ def _calculate_cexs_elem_mat(this, types, temperature=294., # Expand elements in to nuclides with atomic densities nuc_fractions = this.get_nuclide_atom_densities() # Create a dict of [nuclide name] = nuclide object to carry forward - # with a common nuclides format between openmc.Material and - # openmc.Element objects + # with a common nuclides format between openmc.Material and Elements nuclides = {nuclide: nuclide for nuclide in nuc_fractions} else: # Expand elements in to nuclides with atomic densities - nuclides = this.expand(1., 'ao', enrichment=enrichment, + nuclides = openmc.Element(this).expand(1., 'ao', enrichment=enrichment, cross_sections=cross_sections) # For ease of processing split out the nuclide and its fraction nuc_fractions = {nuclide[0]: nuclide[1] for nuclide in nuclides} # Create a dict of [nuclide name] = nuclide object to carry forward - # with a common nuclides format between openmc.Material and - # openmc.Element objects + # with a common nuclides format between openmc.Material and Elements nuclides = {nuclide[0]: nuclide[0] for nuclide in nuclides} # Identify the nuclides which have S(a,b) data @@ -615,7 +603,7 @@ def calculate_mgxs(this, types, orders=None, temperature=294., Parameters ---------- - this : {str, openmc.Nuclide, openmc.Element, openmc.Macroscopic, openmc.Material} + this : {str, openmc.Material} Object to source data from. Nuclides and Elements can be input as a str types : Iterable of values of PLOT_TYPES_MGXS The type of cross sections to calculate @@ -632,7 +620,6 @@ def calculate_mgxs(this, types, orders=None, temperature=294., Location of MGXS HDF5 Library file. Default is None. ce_cross_sections : str, optional Location of continuous-energy cross_sections.xml file. Default is None. - This is used only for expanding an openmc.Element object passed as this enrichment : float, optional Enrichment for U235 in weight percent. For example, input 4.95 for 4.95 weight percent enriched U. Default is None @@ -656,21 +643,13 @@ def calculate_mgxs(this, types, orders=None, temperature=294., cv.check_type("cross_sections", cross_sections, str) library = openmc.MGXSLibrary.from_hdf5(cross_sections) - # this is a nuclide or element if it is a string - if isinstance(this, str): - # first entry in ELEMENT_SYMBOL is a neutron, the 1 removes this entry - if this in list(openmc.data.ELEMENT_SYMBOL.values())[1:]: - this = openmc.Element(this) - else: - this = openmc.Nuclide(this) - - if isinstance(this, openmc.Nuclide) or isinstance(this, openmc.Macroscopic): - mgxs = _calculate_mgxs_nuc_macro(this, types, library, orders, - temperature) - elif isinstance(this, openmc.Element) or isinstance(this, openmc.Material): + if this in ELEMENT_NAMES or isinstance(this, openmc.Material): mgxs = _calculate_mgxs_elem_mat(this, types, library, orders, temperature, ce_cross_sections, enrichment) + elif isinstance(this, str): + mgxs = _calculate_mgxs_nuc_macro(this, types, library, orders, + temperature) else: raise TypeError("Invalid type") @@ -702,7 +681,7 @@ def _calculate_mgxs_nuc_macro(this, types, library, orders=None, Parameters ---------- - this : openmc.Nuclide or openmc.Macroscopic + this : str Object to source data from types : Iterable of str The type of cross sections to calculate; values can either be those @@ -840,8 +819,8 @@ def _calculate_mgxs_elem_mat(this, types, library, orders=None, Parameters ---------- - this : openmc.Element or openmc.Material - Object to source data from + this : str or openmc.Material + Object to source data from. Elements can be input as a str types : Iterable of str The type of cross sections to calculate; values can either be those in openmc.PLOT_TYPES_MGXS @@ -890,7 +869,7 @@ def _calculate_mgxs_elem_mat(this, types, library, orders=None, else: T = temperature # Expand elements in to nuclides with atomic densities - nuclides = this.expand(100., 'ao', enrichment=enrichment, + nuclides = openmc.Element(this).expand(100., 'ao', enrichment=enrichment, cross_sections=ce_cross_sections) # For ease of processing split out nuc and nuc_fractions diff --git a/tests/unit_tests/test_plotter.py b/tests/unit_tests/test_plotter.py index f7c731f1dd..cabf9af6d6 100644 --- a/tests/unit_tests/test_plotter.py +++ b/tests/unit_tests/test_plotter.py @@ -32,7 +32,7 @@ def test_calculate_cexs_elem_mat_sab(test_mat): assert len(data[0]) == len(energy_grid) -@pytest.mark.parametrize("this", ["Li", "Li6", openmc.Nuclide('Li6'), openmc.Element('Li')]) +@pytest.mark.parametrize("this", ["Li", "Li6"]) def test_calculate_cexs_with_nuclide_and_element(this): # single type (reaction) energy_grid, data = openmc.plotter.calculate_cexs( @@ -72,7 +72,7 @@ def test_calculate_cexs_with_materials(test_mat): assert len(data[0]) == len(energy_grid) -@pytest.mark.parametrize("this", ["Be", "Be9", openmc.Nuclide('Be9'), openmc.Element('Be')]) +@pytest.mark.parametrize("this", ["Be", "Be9"]) def test_plot_xs(this): assert isinstance(openmc.plotter.plot_xs(this, types=['total']), Figure)