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merging with upstream
This commit is contained in:
Adam Nelson
commit
ea8181f2b6
57 changed files with 3543 additions and 710 deletions
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@ -1,4 +1,5 @@
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from numbers import Integral, Real
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from six import string_types
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import numpy as np
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@ -19,42 +20,18 @@ UNITY_MT = -1
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XI_MT = -2
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# MTs to combine to generate associated plot_types
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PLOT_TYPES_MT = {'total': (2, 3,),
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'scatter': (2, 4, 11, 16, 17, 22, 23, 24, 25, 28, 29, 30,
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32, 33, 34, 35, 36, 37, 41, 42, 44, 45, 152,
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153, 154, 156, 157, 158, 159, 160, 161, 162,
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163, 164, 165, 166, 167, 168, 169, 170, 171,
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172, 173, 174, 175, 176, 177, 178, 179, 180,
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181, 183, 184, 190, 194, 196, 198, 199, 200,
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875, 891),
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'elastic': (2,),
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'inelastic': (4, 11, 16, 17, 22, 23, 24, 25, 28, 29, 30,
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32, 33, 34, 35, 36, 37, 41, 42, 44, 45, 152,
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153, 154, 156, 157, 158, 159, 160, 161, 162,
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163, 164, 165, 166, 167, 168, 169, 170, 171,
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172, 173, 174, 175, 176, 177, 178, 179, 180,
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181, 183, 184, 190, 194, 196, 198, 199, 200,
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875, 891),
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'fission': (18,),
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'absorption': (27,), 'capture': (101,),
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'nu-fission': (18,),
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'nu-scatter': (2, 4, 11, 16, 17, 22, 23, 24, 25, 28, 29, 30,
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32, 33, 34, 35, 36, 37, 41, 42, 44, 45, 152,
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153, 154, 156, 157, 158, 159, 160, 161, 162,
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163, 164, 165, 166, 167, 168, 169, 170, 171,
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172, 173, 174, 175, 176, 177, 178, 179, 180,
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181, 183, 184, 190, 194, 196, 198, 199, 200,
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875, 891),
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'unity': (UNITY_MT,),
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'slowing-down power': (2, 4, 11, 16, 17, 22, 23, 24, 25, 28,
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29, 30, 32, 33, 34, 35, 36, 37, 41, 42,
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44, 45, 152, 153, 154, 156, 157, 158,
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159, 160, 161, 162, 163, 164, 165, 166,
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167, 168, 169, 170, 171, 172, 173, 174,
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175, 176, 177, 178, 179, 180, 181, 183,
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184, 190, 194, 196, 198, 199, 200, 875,
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891, XI_MT),
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'damage': (444,)}
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_INELASTIC = [mt for mt in openmc.data.SUM_RULES[3] if mt != 27]
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PLOT_TYPES_MT = {'total': openmc.data.SUM_RULES[1],
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'scatter': [2] + _INELASTIC,
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'elastic': [2],
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'inelastic': _INELASTIC,
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'fission': [18],
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'absorption': [27], 'capture': [101],
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'nu-fission': [18],
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'nu-scatter': [2] + _INELASTIC,
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'unity': [UNITY_MT],
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'slowing-down power': [2] + _INELASTIC + [XI_MT],
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'damage': [444]}
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# Operations to use when combining MTs the first np.add is used in reference
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# to zero
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PLOT_TYPES_OP = {'total': (np.add,),
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@ -69,27 +46,13 @@ PLOT_TYPES_OP = {'total': (np.add,),
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(np.add,) * (len(PLOT_TYPES_MT['slowing-down power']) - 2) + (np.multiply,),
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'damage': ()}
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# Whether or not to multiply the reaction by the yield as well
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PLOT_TYPES_YIELD = {'total': (False, False),
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'scatter': (False,) * len(PLOT_TYPES_MT['scatter']),
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'elastic': (False,),
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'inelastic': (False,) * len(PLOT_TYPES_MT['inelastic']),
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'fission': (False,), 'absorption': (False,),
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'capture': (False,), 'nu-fission': (True,),
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'nu-scatter': (True,) * len(PLOT_TYPES_MT['nu-scatter']),
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'unity': (False,),
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'slowing-down power':
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(True,) * len(PLOT_TYPES_MT['slowing-down power']),
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'damage': (False,)}
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# Types of plots to plot linearly in y
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PLOT_TYPES_LINEAR = {'nu-fission / fission', 'nu-scatter / scatter',
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'nu-fission / absorption', 'fission / absorption'}
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def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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energy_range=(1.E-5, 20.E6), sab_name=None, cross_sections=None,
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enrichment=None, **kwargs):
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sab_name=None, cross_sections=None, enrichment=None, **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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@ -110,8 +73,6 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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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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energy_range : tuple of floats
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Energy range (in eV) to plot the cross section within
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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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@ -196,6 +157,7 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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to_plot = data[i](E)
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else:
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to_plot = data[i, :]
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to_plot = np.nan_to_num(to_plot)
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if np.sum(to_plot) > 0.:
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plot_func(E, to_plot, label=types[i])
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@ -216,10 +178,10 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
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ylabel = 'Macroscopic Cross Section [1/cm]'
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ax.set_ylabel(ylabel)
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ax.legend(loc='best')
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ax.set_xlim(energy_range)
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# Set to the most likely expected range
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ax.set_xlim((1.E-5, 20.E6))
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if this.name is not None:
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title = 'Cross Section for ' + this.name
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ax.set_title(title)
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ax.set_title('Cross Section for ' + this.name)
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return fig
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@ -384,7 +346,7 @@ def calculate_xs(this, types, temperature=294., sab_name=None,
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Returns
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-------
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energy_grid : numpy.array
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energy_grid : numpy.ndarray
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Energies at which cross sections are calculated, in units of eV
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data : numpy.ndarray
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Cross sections calculated at the energy grid described by energy_grid
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@ -394,7 +356,7 @@ def calculate_xs(this, types, temperature=294., sab_name=None,
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# Check types
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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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cv.check_type('sab_name', sab_name, string_types)
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if enrichment:
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cv.check_type('enrichment', enrichment, Real)
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@ -440,7 +402,7 @@ def _calculate_xs_element(this, types, temperature=294., sab_name=None,
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Returns
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-------
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energy_grid : numpy.array
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energy_grid : numpy.ndarray
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Energies at which cross sections are calculated, in units of eV
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data : numpy.ndarray
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Macroscopic cross sections calculated at the energy grid described
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@ -452,7 +414,7 @@ def _calculate_xs_element(this, types, temperature=294., sab_name=None,
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library = openmc.data.DataLibrary.from_xml(cross_sections)
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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 = this.expand(1., 'ao', enrichment=enrichment,
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cross_sections=cross_sections)
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# For ease of processing split out nuc and nuc_density
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@ -519,7 +481,7 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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Returns
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-------
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energy_grid : numpy.array
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energy_grid : numpy.ndarray
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Energies at which cross sections are calculated, in units of eV
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data : numpy.ndarray
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Cross sections calculated at the energy grid described by
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@ -534,15 +496,18 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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for line in types:
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if line in PLOT_TYPES:
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mts.append(PLOT_TYPES_MT[line])
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yields.append(PLOT_TYPES_YIELD[line])
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if line.startswith('nu'):
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yields.append(True)
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else:
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yields.append(False)
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ops.append(PLOT_TYPES_OP[line])
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else:
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# Not a built-in type, we have to parse it ourselves
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cv.check_type('MT in types', line, Integral)
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cv.check_greater_than('MT in types', line, 0)
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mts.append((line,))
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yields.append((False,))
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ops.append(())
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yields.append(False)
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# Load the library
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library = openmc.data.DataLibrary.from_xml(cross_sections)
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@ -565,7 +530,7 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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for t in range(len(data_Ts)):
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# Take off the "K" and convert to a float
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data_Ts[t] = float(data_Ts[t][:-1])
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min_delta = np.finfo(np.float64).max
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min_delta = float('inf')
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closest_t = -1
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for t in data_Ts:
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if abs(data_Ts[t] - T) < min_delta:
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@ -620,7 +585,7 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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# Get the reaction xs data from the nuclide
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funcs = []
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op = ops[i]
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for mt, yield_check in zip(mt_set, yields[i]):
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for mt in mt_set:
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if mt == 2:
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if sab_name:
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# Then we need to do a piece-wise function of
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@ -632,29 +597,36 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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funcs.append(pw_funcs)
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else:
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funcs.append(nuc[mt].xs[nucT])
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elif mt == 5 and mt in nuc:
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# Only consider the (n,misc) products if neutrons are
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# included in the outgoing channel since (n,misc) is only
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# explicitly needed for scatter cross sections.
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for prod in nuc[mt].products:
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if prod.particle == 'neutron' and \
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prod.emission_mode in ('total', 'prompt'):
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if yields[i]:
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func = openmc.data.Combination(
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[nuc[mt].xs[nucT], prod.yield_],
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[np.multiply])
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else:
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func = nuc[mt].xs[nucT]
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funcs.append(func)
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break
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else:
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funcs.append(lambda x: 0.)
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elif mt in nuc:
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if yield_check:
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found_it = False
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if yields[i]:
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for prod in nuc[mt].products:
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if prod.particle == 'neutron' and \
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prod.emission_mode == 'total':
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prod.emission_mode in ('total', 'prompt'):
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func = openmc.data.Combination(
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[nuc[mt].xs[nucT], prod.yield_],
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[np.multiply])
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funcs.append(func)
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found_it = True
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break
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if not found_it:
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for prod in nuc[mt].products:
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if prod.particle == 'neutron' and \
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prod.emission_mode == 'prompt':
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func = openmc.data.Combination(
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[nuc[mt].xs[nucT],
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prod.yield_], [np.multiply])
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funcs.append(func)
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found_it = True
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break
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if not found_it:
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else:
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# Assume the yield is 1
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funcs.append(nuc[mt].xs[nucT])
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else:
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@ -695,7 +667,7 @@ def _calculate_xs_material(this, types, temperature=294., cross_sections=None):
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Returns
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-------
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energy_grid : numpy.array
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energy_grid : numpy.ndarray
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Energies at which cross sections are calculated, in units of eV
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data : numpy.ndarray
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Macroscopic cross sections calculated at the energy grid described
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@ -712,7 +684,7 @@ def _calculate_xs_material(this, types, temperature=294., cross_sections=None):
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library = openmc.data.DataLibrary.from_xml(cross_sections)
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# Expand elements in to nuclides with atomic densities
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nuclides = this.get_nuclide_atom_densities(cross_sections)
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nuclides = this.get_nuclide_atom_densities()
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# For ease of processing split out nuc and nuc_density
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nuc_densities = [nuclide[1][1] for nuclide in nuclides.items()]
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@ -740,8 +712,8 @@ def _calculate_xs_material(this, types, temperature=294., cross_sections=None):
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# Condense the data for every nuclide
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# First create a union energy grid
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energy_grid = E[0]
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for n in range(1, len(E)):
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energy_grid = np.union1d(energy_grid, E[n])
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for grid in E[1:]:
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energy_grid = np.union1d(energy_grid, grid)
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# Now we can combine all the nuclidic data
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data = np.zeros((len(types), len(energy_grid)))
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