from collections import OrderedDict, defaultdict, namedtuple, Counter
from collections.abc import Iterable
from copy import deepcopy
from numbers import Real
from pathlib import Path
import re
import warnings
from xml.etree import ElementTree as ET
import numpy as np
import openmc
import openmc.data
import openmc.checkvalue as cv
from ._xml import clean_indentation, reorder_attributes
from .mixin import IDManagerMixin
# Units for density supported by OpenMC
DENSITY_UNITS = ('g/cm3', 'g/cc', 'kg/m3', 'atom/b-cm', 'atom/cm3', 'sum',
'macro')
NuclideTuple = namedtuple('NuclideTuple', ['name', 'percent', 'percent_type'])
[docs]class Material(IDManagerMixin):
"""A material composed of a collection of nuclides/elements.
To create a material, one should create an instance of this class, add
nuclides or elements with :meth:`Material.add_nuclide` or
`Material.add_element`, respectively, and set the total material density
with `Material.set_density()`. The material can then be assigned to a cell
using the :attr:`Cell.fill` attribute.
Parameters
----------
material_id : int, optional
Unique identifier for the material. If not specified, an identifier will
automatically be assigned.
name : str, optional
Name of the material. If not specified, the name will be the empty
string.
temperature : float, optional
Temperature of the material in Kelvin. If not specified, the material
inherits the default temperature applied to the model.
Attributes
----------
id : int
Unique identifier for the material
temperature : float
Temperature of the material in Kelvin.
density : float
Density of the material (units defined separately)
density_units : str
Units used for `density`. Can be one of 'g/cm3', 'g/cc', 'kg/m3',
'atom/b-cm', 'atom/cm3', 'sum', or 'macro'. The 'macro' unit only
applies in the case of a multi-group calculation.
depletable : bool
Indicate whether the material is depletable.
nuclides : list of namedtuple
List in which each item is a namedtuple consisting of a nuclide string,
the percent density, and the percent type ('ao' or 'wo'). The namedtuple
has field names ``name``, ``percent``, and ``percent_type``.
isotropic : list of str
Nuclides for which elastic scattering should be treated as though it
were isotropic in the laboratory system.
average_molar_mass : float
The average molar mass of nuclides in the material in units of grams per
mol. For example, UO2 with 3 nuclides will have an average molar mass
of 270 / 3 = 90 g / mol.
volume : float
Volume of the material in cm^3. This can either be set manually or
calculated in a stochastic volume calculation and added via the
:meth:`Material.add_volume_information` method.
paths : list of str
The paths traversed through the CSG tree to reach each material
instance. This property is initialized by calling the
:meth:`Geometry.determine_paths` method.
num_instances : int
The number of instances of this material throughout the geometry. This
property is initialized by calling the :meth:`Geometry.determine_paths`
method.
fissionable_mass : float
Mass of fissionable nuclides in the material in [g]. Requires that the
:attr:`volume` attribute is set.
"""
next_id = 1
used_ids = set()
def __init__(self, material_id=None, name='', temperature=None):
# Initialize class attributes
self.id = material_id
self.name = name
self.temperature = temperature
self._density = None
self._density_units = 'sum'
self._depletable = False
self._paths = None
self._num_instances = None
self._volume = None
self._atoms = {}
self._isotropic = []
# A list of tuples (nuclide, percent, percent type)
self._nuclides = []
# The single instance of Macroscopic data present in this material
# (only one is allowed, hence this is different than _nuclides, etc)
self._macroscopic = None
# If specified, a list of table names
self._sab = []
def __repr__(self):
string = 'Material\n'
string += '{: <16}=\t{}\n'.format('\tID', self._id)
string += '{: <16}=\t{}\n'.format('\tName', self._name)
string += '{: <16}=\t{}\n'.format('\tTemperature', self._temperature)
string += '{: <16}=\t{}'.format('\tDensity', self._density)
string += ' [{}]\n'.format(self._density_units)
string += '{: <16}\n'.format('\tS(a,b) Tables')
for sab in self._sab:
string += '{: <16}=\t{}\n'.format('\tS(a,b)', sab)
string += '{: <16}\n'.format('\tNuclides')
for nuclide, percent, percent_type in self._nuclides:
string += '{: <16}'.format('\t{}'.format(nuclide))
string += '=\t{: <12} [{}]\n'.format(percent, percent_type)
if self._macroscopic is not None:
string += '{: <16}\n'.format('\tMacroscopic Data')
string += '{: <16}'.format('\t{}'.format(self._macroscopic))
return string
@property
def name(self):
return self._name
@property
def temperature(self):
return self._temperature
@property
def density(self):
return self._density
@property
def density_units(self):
return self._density_units
@property
def depletable(self):
return self._depletable
@property
def paths(self):
if self._paths is None:
raise ValueError('Material instance paths have not been determined. '
'Call the Geometry.determine_paths() method.')
return self._paths
@property
def num_instances(self):
if self._num_instances is None:
raise ValueError(
'Number of material instances have not been determined. Call '
'the Geometry.determine_paths() method.')
return self._num_instances
@property
def nuclides(self):
return self._nuclides
@property
def isotropic(self):
return self._isotropic
@property
def average_molar_mass(self):
# Using the sum of specified atomic or weight amounts as a basis, sum
# the mass and moles of the material
mass = 0.
moles = 0.
for nuc in self.nuclides:
if nuc.percent_type == 'ao':
mass += nuc.percent * openmc.data.atomic_mass(nuc.name)
moles += nuc.percent
else:
moles += nuc.percent / openmc.data.atomic_mass(nuc.name)
mass += nuc.percent
# Compute and return the molar mass
return mass / moles
@property
def volume(self):
return self._volume
@name.setter
def name(self, name):
if name is not None:
cv.check_type('name for Material ID="{}"'.format(self._id),
name, str)
self._name = name
else:
self._name = ''
@temperature.setter
def temperature(self, temperature):
cv.check_type('Temperature for Material ID="{}"'.format(self._id),
temperature, (Real, type(None)))
self._temperature = temperature
@depletable.setter
def depletable(self, depletable):
cv.check_type('Depletable flag for Material ID="{}"'.format(self.id),
depletable, bool)
self._depletable = depletable
@volume.setter
def volume(self, volume):
if volume is not None:
cv.check_type('material volume', volume, Real)
self._volume = volume
@isotropic.setter
def isotropic(self, isotropic):
cv.check_iterable_type('Isotropic scattering nuclides', isotropic,
str)
self._isotropic = list(isotropic)
@property
def fissionable_mass(self):
if self.volume is None:
raise ValueError("Volume must be set in order to determine mass.")
density = 0.0
for nuc, atoms_per_cc in self.get_nuclide_atom_densities().values():
Z = openmc.data.zam(nuc)[0]
if Z >= 90:
density += 1e24 * atoms_per_cc * openmc.data.atomic_mass(nuc) \
/ openmc.data.AVOGADRO
return density*self.volume
[docs] @classmethod
def from_hdf5(cls, group):
"""Create material from HDF5 group
Parameters
----------
group : h5py.Group
Group in HDF5 file
Returns
-------
openmc.Material
Material instance
"""
mat_id = int(group.name.split('/')[-1].lstrip('material '))
name = group['name'][()].decode() if 'name' in group else ''
density = group['atom_density'][()]
if 'nuclide_densities' in group:
nuc_densities = group['nuclide_densities'][()]
# Create the Material
material = cls(mat_id, name)
material.depletable = bool(group.attrs['depletable'])
if 'volume' in group.attrs:
material.volume = group.attrs['volume']
if "temperature" in group.attrs:
material.temperature = group.attrs["temperature"]
# Read the names of the S(a,b) tables for this Material and add them
if 'sab_names' in group:
sab_tables = group['sab_names'][()]
for sab_table in sab_tables:
name = sab_table.decode()
material.add_s_alpha_beta(name)
# Set the Material's density to atom/b-cm as used by OpenMC
material.set_density(density=density, units='atom/b-cm')
if 'nuclides' in group:
nuclides = group['nuclides'][()]
# Add all nuclides to the Material
for fullname, density in zip(nuclides, nuc_densities):
name = fullname.decode().strip()
material.add_nuclide(name, percent=density, percent_type='ao')
if 'macroscopics' in group:
macroscopics = group['macroscopics'][()]
# Add all macroscopics to the Material
for fullname in macroscopics:
name = fullname.decode().strip()
material.add_macroscopic(name)
return material
[docs] def set_density(self, units, density=None):
"""Set the density of the material
Parameters
----------
units : {'g/cm3', 'g/cc', 'kg/m3', 'atom/b-cm', 'atom/cm3', 'sum', 'macro'}
Physical units of density.
density : float, optional
Value of the density. Must be specified unless units is given as
'sum'.
"""
cv.check_value('density units', units, DENSITY_UNITS)
self._density_units = units
if units == 'sum':
if density is not None:
msg = 'Density "{}" for Material ID="{}" is ignored ' \
'because the unit is "sum"'.format(density, self.id)
warnings.warn(msg)
else:
if density is None:
msg = 'Unable to set the density for Material ID="{}" ' \
'because a density value must be given when not using ' \
'"sum" unit'.format(self.id)
raise ValueError(msg)
cv.check_type('the density for Material ID="{}"'.format(self.id),
density, Real)
self._density = density
[docs] def add_nuclide(self, nuclide, percent, percent_type='ao'):
"""Add a nuclide to the material
Parameters
----------
nuclide : str
Nuclide to add, e.g., 'Mo95'
percent : float
Atom or weight percent
percent_type : {'ao', 'wo'}
'ao' for atom percent and 'wo' for weight percent
"""
cv.check_type('nuclide', nuclide, str)
cv.check_type('percent', percent, Real)
cv.check_value('percent type', percent_type, {'ao', 'wo'})
if self._macroscopic is not None:
msg = 'Unable to add a Nuclide to Material ID="{}" as a ' \
'macroscopic data-set has already been added'.format(self._id)
raise ValueError(msg)
# If nuclide name doesn't look valid, give a warning
try:
Z, _, _ = openmc.data.zam(nuclide)
except ValueError as e:
warnings.warn(str(e))
else:
# For actinides, have the material be depletable by default
if Z >= 89:
self.depletable = True
self._nuclides.append(NuclideTuple(nuclide, percent, percent_type))
[docs] def remove_nuclide(self, nuclide):
"""Remove a nuclide from the material
Parameters
----------
nuclide : str
Nuclide to remove
"""
cv.check_type('nuclide', nuclide, str)
# If the Material contains the Nuclide, delete it
for nuc in reversed(self.nuclides):
if nuclide == nuc.name:
self.nuclides.remove(nuc)
[docs] def add_macroscopic(self, macroscopic):
"""Add a macroscopic to the material. This will also set the
density of the material to 1.0, unless it has been otherwise set,
as a default for Macroscopic cross sections.
Parameters
----------
macroscopic : str
Macroscopic to add
"""
# Ensure no nuclides, elements, or sab are added since these would be
# incompatible with macroscopics
if self._nuclides or self._sab:
msg = 'Unable to add a Macroscopic data set to Material ID="{}" ' \
'with a macroscopic value "{}" as an incompatible data ' \
'member (i.e., nuclide or S(a,b) table) ' \
'has already been added'.format(self._id, macroscopic)
raise ValueError(msg)
if not isinstance(macroscopic, str):
msg = 'Unable to add a Macroscopic to Material ID="{}" with a ' \
'non-string value "{}"'.format(self._id, macroscopic)
raise ValueError(msg)
if self._macroscopic is None:
self._macroscopic = macroscopic
else:
msg = 'Unable to add a Macroscopic to Material ID="{}". ' \
'Only one Macroscopic allowed per ' \
'Material.'.format(self._id)
raise ValueError(msg)
# Generally speaking, the density for a macroscopic object will
# be 1.0. Therefore, lets set density to 1.0 so that the user
# doesnt need to set it unless its needed.
# Of course, if the user has already set a value of density,
# then we will not override it.
if self._density is None:
self.set_density('macro', 1.0)
[docs] def remove_macroscopic(self, macroscopic):
"""Remove a macroscopic from the material
Parameters
----------
macroscopic : str
Macroscopic to remove
"""
if not isinstance(macroscopic, str):
msg = 'Unable to remove a Macroscopic "{}" in Material ID="{}" ' \
'since it is not a string'.format(self._id, macroscopic)
raise ValueError(msg)
# If the Material contains the Macroscopic, delete it
if macroscopic == self._macroscopic:
self._macroscopic = None
[docs] def add_element(self, element, percent, percent_type='ao', enrichment=None,
enrichment_target=None, enrichment_type=None):
"""Add a natural element to the material
Parameters
----------
element : str
Element to add, e.g., 'Zr' or 'Zirconium'
percent : float
Atom or weight percent
percent_type : {'ao', 'wo'}, optional
'ao' for atom percent and 'wo' for weight percent. Defaults to atom
percent.
enrichment : float, optional
Enrichment of an enrichment_taget nuclide in percent (ao or wo).
If enrichment_taget is not supplied then it is enrichment for U235
in weight percent. For example, input 4.95 for 4.95 weight percent
enriched U.
Default is None (natural composition).
enrichment_target: str, optional
Single nuclide name to enrich from a natural composition (e.g., 'O16')
.. versionadded:: 0.12
enrichment_type: {'ao', 'wo'}, optional
'ao' for enrichment as atom percent and 'wo' for weight percent.
Default is: 'ao' for two-isotope enrichment; 'wo' for U enrichment
.. versionadded:: 0.12
Notes
-----
General enrichment procedure is allowed only for elements composed of
two isotopes. If `enrichment_target` is given without `enrichment`
natural composition is added to the material.
"""
cv.check_type('nuclide', element, str)
cv.check_type('percent', percent, Real)
cv.check_value('percent type', percent_type, {'ao', 'wo'})
# Make sure element name is just that
if not element.isalpha():
raise ValueError("Element name should be given by the "
"element's symbol or name, e.g., 'Zr', 'zirconium'")
# Allow for element identifier to be given as a symbol or name
if len(element) > 2:
el = element.lower()
element = openmc.data.ELEMENT_SYMBOL.get(el)
if element is None:
msg = 'Element name "{}" not recognised'.format(el)
raise ValueError(msg)
if self._macroscopic is not None:
msg = 'Unable to add an Element to Material ID="{}" as a ' \
'macroscopic data-set has already been added'.format(self._id)
raise ValueError(msg)
if enrichment is not None and enrichment_target is None:
if not isinstance(enrichment, Real):
msg = 'Unable to add an Element to Material ID="{}" with a ' \
'non-floating point enrichment value "{}"'\
.format(self._id, enrichment)
raise ValueError(msg)
elif element != 'U':
msg = 'Unable to use enrichment for element {} which is not ' \
'uranium for Material ID="{}"'.format(element, self._id)
raise ValueError(msg)
# Check that the enrichment is in the valid range
cv.check_less_than('enrichment', enrichment, 100./1.008)
cv.check_greater_than('enrichment', enrichment, 0., equality=True)
if enrichment > 5.0:
msg = 'A uranium enrichment of {} was given for Material ID='\
'"{}". OpenMC assumes the U234/U235 mass ratio is '\
'constant at 0.008, which is only valid at low ' \
'enrichments. Consider setting the isotopic ' \
'composition manually for enrichments over 5%.'.\
format(enrichment, self._id)
warnings.warn(msg)
# Add naturally-occuring isotopes
element = openmc.Element(element)
for nuclide in element.expand(percent,
percent_type,
enrichment,
enrichment_target,
enrichment_type):
self.add_nuclide(*nuclide)
[docs] def add_s_alpha_beta(self, name, fraction=1.0):
r"""Add an :math:`S(\alpha,\beta)` table to the material
Parameters
----------
name : str
Name of the :math:`S(\alpha,\beta)` table
fraction : float
The fraction of relevant nuclei that are affected by the
:math:`S(\alpha,\beta)` table. For example, if the material is a
block of carbon that is 60% graphite and 40% amorphous then add a
graphite :math:`S(\alpha,\beta)` table with fraction=0.6.
"""
if self._macroscopic is not None:
msg = 'Unable to add an S(a,b) table to Material ID="{}" as a ' \
'macroscopic data-set has already been added'.format(self._id)
raise ValueError(msg)
if not isinstance(name, str):
msg = 'Unable to add an S(a,b) table to Material ID="{}" with a ' \
'non-string table name "{}"'.format(self._id, name)
raise ValueError(msg)
cv.check_type('S(a,b) fraction', fraction, Real)
cv.check_greater_than('S(a,b) fraction', fraction, 0.0, True)
cv.check_less_than('S(a,b) fraction', fraction, 1.0, True)
new_name = openmc.data.get_thermal_name(name)
if new_name != name:
msg = 'OpenMC S(a,b) tables follow the GND naming convention. ' \
'Table "{}" is being renamed as "{}".'.format(name, new_name)
warnings.warn(msg)
self._sab.append((new_name, fraction))
def make_isotropic_in_lab(self):
self.isotropic = [x.name for x in self._nuclides]
[docs] def get_elements(self):
"""Returns all elements in the material
.. versionadded:: 0.12
Returns
-------
elements : list of str
List of element names
"""
return sorted({re.split(r'(\d+)', i)[0] for i in self.get_nuclides()})
[docs] def get_nuclides(self):
"""Returns all nuclides in the material
Returns
-------
nuclides : list of str
List of nuclide names
"""
return [x.name for x in self._nuclides]
[docs] def get_nuclide_densities(self):
"""Returns all nuclides in the material and their densities
Returns
-------
nuclides : dict
Dictionary whose keys are nuclide names and values are 3-tuples of
(nuclide, density percent, density percent type)
"""
# keep ordered dictionary for testing purposes
nuclides = OrderedDict()
for nuclide in self._nuclides:
nuclides[nuclide.name] = nuclide
return nuclides
[docs] def get_nuclide_atom_densities(self):
"""Returns all nuclides in the material and their atomic densities in
units of atom/b-cm
Returns
-------
nuclides : dict
Dictionary whose keys are nuclide names and values are tuples of
(nuclide, density in atom/b-cm)
"""
sum_density = False
if self.density_units == 'sum':
sum_density = True
density = 0.
elif self.density_units == 'macro':
density = self.density
elif self.density_units == 'g/cc' or self.density_units == 'g/cm3':
density = -self.density
elif self.density_units == 'kg/m3':
density = -0.001 * self.density
elif self.density_units == 'atom/b-cm':
density = self.density
elif self.density_units == 'atom/cm3' or self.density_units == 'atom/cc':
density = 1.E-24 * self.density
# For ease of processing split out nuc, nuc_density,
# and nuc_density_type into separate arrays
nucs = []
nuc_densities = []
nuc_density_types = []
for nuclide in self.nuclides:
nucs.append(nuclide.name)
nuc_densities.append(nuclide.percent)
nuc_density_types.append(nuclide.percent_type)
nucs = np.array(nucs)
nuc_densities = np.array(nuc_densities)
nuc_density_types = np.array(nuc_density_types)
if sum_density:
density = np.sum(nuc_densities)
percent_in_atom = np.all(nuc_density_types == 'ao')
density_in_atom = density > 0.
sum_percent = 0.
# Convert the weight amounts to atomic amounts
if not percent_in_atom:
for n, nuc in enumerate(nucs):
nuc_densities[n] *= self.average_molar_mass / \
openmc.data.atomic_mass(nuc)
# Now that we have the atomic amounts, lets finish calculating densities
sum_percent = np.sum(nuc_densities)
nuc_densities = nuc_densities / sum_percent
# Convert the mass density to an atom density
if not density_in_atom:
density = -density / self.average_molar_mass * 1.E-24 \
* openmc.data.AVOGADRO
nuc_densities = density * nuc_densities
nuclides = OrderedDict()
for n, nuc in enumerate(nucs):
nuclides[nuc] = (nuc, nuc_densities[n])
return nuclides
[docs] def get_mass_density(self, nuclide=None):
"""Return mass density of one or all nuclides
Parameters
----------
nuclides : str, optional
Nuclide for which density is desired. If not specified, the density
for the entire material is given.
Returns
-------
float
Density of the nuclide/material in [g/cm^3]
"""
mass_density = 0.0
for nuc, atoms_per_cc in self.get_nuclide_atom_densities().values():
if nuclide is None or nuclide == nuc:
density_i = 1e24 * atoms_per_cc * openmc.data.atomic_mass(nuc) \
/ openmc.data.AVOGADRO
mass_density += density_i
return mass_density
[docs] def get_mass(self, nuclide=None):
"""Return mass of one or all nuclides.
Note that this method requires that the :attr:`Material.volume` has
already been set.
Parameters
----------
nuclides : str, optional
Nuclide for which mass is desired. If not specified, the density
for the entire material is given.
Returns
-------
float
Mass of the nuclide/material in [g]
"""
if self.volume is None:
raise ValueError("Volume must be set in order to determine mass.")
return self.volume*self.get_mass_density(nuclide)
[docs] def clone(self, memo=None):
"""Create a copy of this material with a new unique ID.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Material
The clone of this material
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
# Temporarily remove paths -- this is done so that when the clone is
# made, it doesn't create a copy of the paths (which are specific to
# an instance)
paths = self._paths
self._paths = None
clone = deepcopy(self)
clone.id = None
clone._num_instances = None
# Restore paths on original instance
self._paths = paths
# Memoize the clone
memo[self] = clone
return memo[self]
def _get_nuclide_xml(self, nuclide):
xml_element = ET.Element("nuclide")
xml_element.set("name", nuclide.name)
if nuclide.percent_type == 'ao':
xml_element.set("ao", str(nuclide.percent))
else:
xml_element.set("wo", str(nuclide.percent))
return xml_element
def _get_macroscopic_xml(self, macroscopic):
xml_element = ET.Element("macroscopic")
xml_element.set("name", macroscopic)
return xml_element
def _get_nuclides_xml(self, nuclides):
xml_elements = []
for nuclide in nuclides:
xml_elements.append(self._get_nuclide_xml(nuclide))
return xml_elements
[docs] def to_xml_element(self, cross_sections=None):
"""Return XML representation of the material
Parameters
----------
cross_sections : str
Path to an XML cross sections listing file
Returns
-------
element : xml.etree.ElementTree.Element
XML element containing material data
"""
# Create Material XML element
element = ET.Element("material")
element.set("id", str(self._id))
if len(self._name) > 0:
element.set("name", str(self._name))
if self._depletable:
element.set("depletable", "true")
if self._volume:
element.set("volume", str(self._volume))
# Create temperature XML subelement
if self.temperature is not None:
element.set("temperature", str(self.temperature))
# Create density XML subelement
if self._density is not None or self._density_units == 'sum':
subelement = ET.SubElement(element, "density")
if self._density_units != 'sum':
subelement.set("value", str(self._density))
subelement.set("units", self._density_units)
else:
raise ValueError('Density has not been set for material {}!'
.format(self.id))
if self._macroscopic is None:
# Create nuclide XML subelements
subelements = self._get_nuclides_xml(self._nuclides)
for subelement in subelements:
element.append(subelement)
else:
# Create macroscopic XML subelements
subelement = self._get_macroscopic_xml(self._macroscopic)
element.append(subelement)
if self._sab:
for sab in self._sab:
subelement = ET.SubElement(element, "sab")
subelement.set("name", sab[0])
if sab[1] != 1.0:
subelement.set("fraction", str(sab[1]))
if self._isotropic:
subelement = ET.SubElement(element, "isotropic")
subelement.text = ' '.join(self._isotropic)
return element
[docs] @classmethod
def mix_materials(cls, materials, fracs, percent_type='ao', name=None):
"""Mix materials together based on atom, weight, or volume fractions
.. versionadded:: 0.12
Parameters
----------
materials : Iterable of openmc.Material
Materials to combine
fracs : Iterable of float
Fractions of each material to be combined
percent_type : {'ao', 'wo', 'vo'}
Type of percentage, must be one of 'ao', 'wo', or 'vo', to signify atom
percent (molar percent), weight percent, or volume percent,
optional. Defaults to 'ao'
name : str
The name for the new material, optional. Defaults to concatenated
names of input materials with percentages indicated inside
parentheses.
Returns
-------
openmc.Material
Mixture of the materials
"""
cv.check_type('materials', materials, Iterable, Material)
cv.check_type('fracs', fracs, Iterable, Real)
cv.check_value('percent type', percent_type, {'ao', 'wo', 'vo'})
fracs = np.asarray(fracs)
void_frac = 1. - np.sum(fracs)
# Warn that fractions don't add to 1, set remainder to void, or raise
# an error if percent_type isn't 'vo'
if not np.isclose(void_frac, 0.):
if percent_type in ('ao', 'wo'):
msg = ('A non-zero void fraction is not acceptable for '
'percent_type: {}'.format(percent_type))
raise ValueError(msg)
else:
msg = ('Warning: sum of fractions do not add to 1, void '
'fraction set to {}'.format(void_frac))
warnings.warn(msg)
# Calculate appropriate weights which are how many cc's of each
# material are found in 1cc of the composite material
amms = np.asarray([mat.average_molar_mass for mat in materials])
mass_dens = np.asarray([mat.get_mass_density() for mat in materials])
if percent_type == 'ao':
wgts = fracs * amms / mass_dens
wgts /= np.sum(wgts)
elif percent_type == 'wo':
wgts = fracs / mass_dens
wgts /= np.sum(wgts)
elif percent_type == 'vo':
wgts = fracs
# If any of the involved materials contain S(a,b) tables raise an error
sab_names = set(sab[0] for mat in materials for sab in mat._sab)
if sab_names:
msg = ('Currently we do not support mixing materials containing '
'S(a,b) tables')
raise NotImplementedError(msg)
# Add nuclide densities weighted by appropriate fractions
nuclides_per_cc = defaultdict(float)
mass_per_cc = defaultdict(float)
for mat, wgt in zip(materials, wgts):
for nuc, atoms_per_bcm in mat.get_nuclide_atom_densities().values():
nuc_per_cc = wgt*1.e24*atoms_per_bcm
nuclides_per_cc[nuc] += nuc_per_cc
mass_per_cc[nuc] += nuc_per_cc*openmc.data.atomic_mass(nuc) / \
openmc.data.AVOGADRO
# Create the new material with the desired name
if name is None:
name = '-'.join(['{}({})'.format(m.name, f) for m, f in
zip(materials, fracs)])
new_mat = openmc.Material(name=name)
# Compute atom fractions of nuclides and add them to the new material
tot_nuclides_per_cc = np.sum([dens for dens in nuclides_per_cc.values()])
for nuc, atom_dens in nuclides_per_cc.items():
new_mat.add_nuclide(nuc, atom_dens/tot_nuclides_per_cc, 'ao')
# Compute mass density for the new material and set it
new_density = np.sum([dens for dens in mass_per_cc.values()])
new_mat.set_density('g/cm3', new_density)
# If any of the involved materials is depletable, the new material is
# depletable
new_mat.depletable = any(mat.depletable for mat in materials)
return new_mat
[docs] @classmethod
def from_xml_element(cls, elem):
"""Generate material from an XML element
Parameters
----------
elem : xml.etree.ElementTree.Element
XML element
Returns
-------
openmc.Material
Material generated from XML element
"""
mat_id = int(elem.get('id'))
mat = cls(mat_id)
mat.name = elem.get('name')
if "temperature" in elem.attrib:
mat.temperature = float(elem.get("temperature"))
if 'volume' in elem.attrib:
mat.volume = float(elem.get('volume'))
mat.depletable = bool(elem.get('depletable'))
# Get each nuclide
for nuclide in elem.findall('nuclide'):
name = nuclide.attrib['name']
if 'ao' in nuclide.attrib:
mat.add_nuclide(name, float(nuclide.attrib['ao']))
elif 'wo' in nuclide.attrib:
mat.add_nuclide(name, float(nuclide.attrib['wo']), 'wo')
# Get each S(a,b) table
for sab in elem.findall('sab'):
fraction = float(sab.get('fraction', 1.0))
mat.add_s_alpha_beta(sab.get('name'), fraction)
# Get total material density
density = elem.find('density')
units = density.get('units')
if units == 'sum':
mat.set_density(units)
else:
value = float(density.get('value'))
mat.set_density(units, value)
# Check for isotropic scattering nuclides
isotropic = elem.find('isotropic')
if isotropic is not None:
mat.isotropic = isotropic.text.split()
return mat
[docs]class Materials(cv.CheckedList):
"""Collection of Materials used for an OpenMC simulation.
This class corresponds directly to the materials.xml input file. It can be
thought of as a normal Python list where each member is a
:class:`Material`. It behaves like a list as the following example
demonstrates:
>>> fuel = openmc.Material()
>>> clad = openmc.Material()
>>> water = openmc.Material()
>>> m = openmc.Materials([fuel])
>>> m.append(water)
>>> m += [clad]
Parameters
----------
materials : Iterable of openmc.Material
Materials to add to the collection
cross_sections : str
Indicates the path to an XML cross section listing file (usually named
cross_sections.xml). If it is not set, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used for
continuous-energy calculations and
:envvar:`OPENMC_MG_CROSS_SECTIONS` will be used for multi-group
calculations to find the path to the HDF5 cross section file.
"""
def __init__(self, materials=None):
super().__init__(Material, 'materials collection')
self._cross_sections = None
if materials is not None:
self += materials
@property
def cross_sections(self):
return self._cross_sections
@cross_sections.setter
def cross_sections(self, cross_sections):
cv.check_type('cross sections', cross_sections, str)
self._cross_sections = cross_sections
[docs] def append(self, material):
"""Append material to collection
Parameters
----------
material : openmc.Material
Material to append
"""
super().append(material)
[docs] def insert(self, index, material):
"""Insert material before index
Parameters
----------
index : int
Index in list
material : openmc.Material
Material to insert
"""
super().insert(index, material)
def make_isotropic_in_lab(self):
for material in self:
material.make_isotropic_in_lab()
[docs] def export_to_xml(self, path='materials.xml'):
"""Export material collection to an XML file.
Parameters
----------
path : str
Path to file to write. Defaults to 'materials.xml'.
"""
# Check if path is a directory
p = Path(path)
if p.is_dir():
p /= 'materials.xml'
# Write materials to the file one-at-a-time. This significantly reduces
# memory demand over allocating a complete ElementTree and writing it in
# one go.
with open(str(p), 'w', encoding='utf-8',
errors='xmlcharrefreplace') as fh:
# Write the header and the opening tag for the root element.
fh.write("<?xml version='1.0' encoding='utf-8'?>\n")
fh.write('<materials>\n')
# Write the <cross_sections> element.
if self._cross_sections is not None:
element = ET.Element('cross_sections')
element.text = str(self._cross_sections)
clean_indentation(element, level=1)
element.tail = element.tail.strip(' ')
fh.write(' ')
reorder_attributes(element) # TODO: Remove when support is Python 3.8+
ET.ElementTree(element).write(fh, encoding='unicode')
# Write the <material> elements.
for material in sorted(self, key=lambda x: x.id):
element = material.to_xml_element(self.cross_sections)
clean_indentation(element, level=1)
element.tail = element.tail.strip(' ')
fh.write(' ')
reorder_attributes(element) # TODO: Remove when support is Python 3.8+
ET.ElementTree(element).write(fh, encoding='unicode')
# Write the closing tag for the root element.
fh.write('</materials>\n')
[docs] @classmethod
def from_xml(cls, path='materials.xml'):
"""Generate materials collection from XML file
Parameters
----------
path : str, optional
Path to materials XML file
Returns
-------
openmc.Materials
Materials collection
"""
tree = ET.parse(path)
root = tree.getroot()
# Generate each material
materials = cls()
for material in root.findall('material'):
materials.append(Material.from_xml_element(material))
# Check for cross sections settings
xs = tree.find('cross_sections')
if xs is not None:
materials.cross_sections = xs.text
return materials