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Resolving @paulromano comments
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Adam Nelson
parent
625903e968
commit
b8d9fe461f
3 changed files with 40 additions and 68 deletions
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@ -186,3 +186,9 @@ K_BOLTZMANN = 8.6173324e-5
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# Used for converting units in ACE data
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EV_PER_MEV = 1.0e6
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# Avogadro's constant from CODATA 2010
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AVOGADRO = 6.02214129E23
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# Neutron mass from CODATA 2010 in units of amu
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NEUTRON_MASS = 1.008664916
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@ -664,17 +664,12 @@ class Material(object):
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sum_percent = 0.
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awrs = []
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n = -1
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for nuclide in nuclides.items():
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n += 1
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lib = library.get_by_material(nuclide[0])
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if lib is not None:
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nuc = openmc.data.IncidentNeutron.from_hdf5(lib['path'])
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awrs.append(nuc.atomic_weight_ratio)
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if not percent_in_atom:
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nuc_densities[n] = -nuc_densities[n] / awrs[-1]
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for n, nuclide in enumerate(nuclides.items()):
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awr = openmc.data.atomic_mass(nuclide[0])
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if awr is not None:
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awrs.append(awr / openmc.data.NEUTRON_MASS)
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else:
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raise ValueError(nuclide[0] + " not in library")
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raise ValueError(nuclide[0] + " is invalid")
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# Now that we have the awr, lets finish calculating densities
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sum_percent = np.sum(nuc_densities)
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@ -686,7 +681,7 @@ class Material(object):
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sum_percent += x * awrs[n]
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sum_percent = 1. / sum_percent
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density = -density * sum_percent * \
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sc.Avogadro / sc.value('neutron mass in u') * 1.E-24
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openmc.data.AVOGADRO / openmc.data.NEUTRON_MASS * 1.E-24
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nuc_densities = density * nuc_densities
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nuclides = OrderedDict()
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@ -15,42 +15,18 @@ 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 not in [5, 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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@ -84,8 +60,7 @@ PLOT_TYPES_LINEAR = {'nu-fission / fission', 'nu-scatter / scatter',
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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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@ -106,8 +81,6 @@ 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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@ -212,10 +185,10 @@ 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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@ -246,7 +219,7 @@ 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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@ -302,7 +275,7 @@ 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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@ -381,7 +354,7 @@ 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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@ -405,6 +378,7 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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mts.append((line,))
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yields.append((False,))
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ops.append(())
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print(mts)
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# Load the library
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library = openmc.data.DataLibrary.from_xml(cross_sections)
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@ -427,7 +401,7 @@ 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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@ -496,7 +470,6 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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funcs.append(nuc[mt].xs[nucT])
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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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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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@ -504,9 +477,8 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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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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else:
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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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@ -514,11 +486,10 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
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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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# Assume the yield is 1
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funcs.append(nuc[mt].xs[nucT])
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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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funcs.append(nuc[mt].xs[nucT])
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elif mt == UNITY_MT:
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@ -557,7 +528,7 @@ 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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@ -602,8 +573,8 @@ 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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