5. Material Compositions¶
Materials in OpenMC are defined as a set of nuclides/elements at specified
densities and are created using the openmc.Material
class. Once a
material has been instantiated, nuclides can be added with
Material.add_nuclide()
and elements can be added with
Material.add_element()
. Densities can be specified using atom fractions or
weight fractions. For example, to create a material and add Gd152 at 0.5 atom
percent, you’d run:
mat = openmc.Material()
mat.add_nuclide('Gd152', 0.5, 'ao')
The third argument to Material.add_nuclide()
can also be ‘wo’ for weight
percent. The densities specified for each nuclide/element are relative and are
renormalized based on the total density of the material. The total density is
set using the Material.set_density()
method. The density can be specified
in gram per cubic centimeter (‘g/cm3’), atom per barn-cm (‘atom/b-cm’), or
kilogram per cubic meter (‘kg/m3’), e.g.,
mat.set_density('g/cm3', 4.5)
5.1. Natural Elements¶
The Material.add_element()
method works exactly the same as
Material.add_nuclide()
, except that instead of specifying a single isotope
of an element, you specify the element itself. For example,
mat.add_element('C', 1.0)
This method can also accept case-insensitive element names such as
mat.add_element('aluminium', 1.0)
Internally, OpenMC stores data on the atomic masses and natural abundances of all known isotopes and then uses this data to determine what isotopes should be added to the material. When the material is later exported to XML for use by the openmc executable, you’ll see that any natural elements were expanded to the naturally-occurring isotopes.
The Material.add_element()
method can also be used to add uranium at a
specified enrichment through the enrichment argument. For example, the
following would add 3.2% enriched uranium to a material:
mat.add_element('U', 1.0, enrichment=3.2)
In addition to U235 and U238, concentrations of U234 and U236 will be present and are determined through a correlation based on measured data.
It is also possible to perform enrichment of any element that is composed
of two naturally-occurring isotopes (e.g., Li or B) in terms of atomic percent.
To invoke this, provide the additional argument enrichment_target to
Material.add_element()
. For example the following would enrich B10
to 30ao%:
mat.add_element('B', 1.0, enrichment=30.0, enrichment_target='B10')
In order to enrich an isotope in terms of mass percent (wo%), provide the extra argument enrichment_type. For example the following would enrich Li6 to 15wo%:
mat.add_element('Li', 1.0, enrichment=15.0, enrichment_target='Li6',
enrichment_type='wo')
Often, cross section libraries don’t actually have all naturally-occurring
isotopes for a given element. For example, in ENDF/B-VII.1, cross section
evaluations are given for O16 and O17 but not for O18. If OpenMC is aware of
what cross sections you will be using (through the
OPENMC_CROSS_SECTIONS
environment variable), it will attempt to only
put isotopes in your model for which you have cross section data. In the case of
oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped with O16.
5.2. Thermal Scattering Data¶
If you have a moderating material in your model like water or graphite, you
should assign thermal scattering data (so-called \(S(\alpha,\beta)\)) using
the Material.add_s_alpha_beta()
method. For example, to model light water,
you would need to add hydrogen and oxygen to a material and then assign the
c_H_in_H2O
thermal scattering data:
water = openmc.Material()
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.add_s_alpha_beta('c_H_in_H2O')
water.set_density('g/cm3', 1.0)
5.3. Adding NCrystal materials¶
Additional support for thermal scattering can be added by using NCrystal. The
Material.from_ncrystal()
class method generates a openmc.Material
object from an NCrystal configuration string.
Temperature, material composition, and density are passed from the configuration
string and the NCMAT file that define the
material, e.g.:
mat = openmc.Material.from_ncrystal('Al_sg225.ncmat;temp=300K')
defines a material containing polycrystalline alumnium,
mat = openmc.Material.from_ncrystal("""Ge_sg227.ncmat;dcutoff=0.5;mos=40arcsec;
dir1=@crys_hkl:5,1,1@lab:0,0,1;
dir2=@crys_hkl:0,-1,1@lab:0,1,0""")
defines an oriented germanium single crystal with 40 arcsec mosaicity.
NCrystal only handles low energy neutron interactions. Other interactions are provided by standard ACE files. NCrystal comes with a predefined library but more materials can be added by creating NCMAT files or on-the-fly in the configuration string.
Warning
Currently, NCrystal materials cannot be modified after they are created. Density, temperature and composition should be defined in the configuration string or the NCMAT file.
5.4. Naming Conventions¶
OpenMC uses the GNDS naming convention for nuclides, metastable states, and compounds:
- Nuclides
SymA
where “A” is the mass number (e.g.,Fe56
)- Elements
Sym0
(e.g.,Fe0
orC0
)- Excited states
SymA_eN
(e.g.,V51_e1
for the first excited state of Vanadium-51.) This is only used in decay data.- Metastable states
SymA_mN
(e.g.,Am242_m1
for the first excited state of Americium-242).- Compounds
c_String_Describing_Material
(e.g.,c_H_in_H2O
). Used for thermal scattering data.
Important
The element syntax, e.g., C0
, is only used when the cross
section evaluation is an elemental evaluation, like carbon in
ENDF/B-VII.1! If you are adding an element via
Material.add_element()
, just use Sym
.
5.5. Temperature¶
Some Monte Carlo codes define temperature implicitly through the cross section data, which is itself given only at a particular temperature. In OpenMC, the material definition is decoupled from the specification of temperature. Instead, temperatures are assigned to cells directly. Alternatively, a default temperature can be assigned to a material that is to be applied to any cell where the material is used. In the absence of any cell or material temperature specification, a global default temperature can be set that is applied to all cells and materials. Anytime a material temperature is specified, it will override the global default temperature. Similarly, anytime a cell temperatures is specified, it will override the material or global default temperature. All temperatures should be given in units of Kelvin.
To assign a default material temperature, one should use the temperature
attribute, e.g.,
hot_fuel = openmc.Material()
hot_fuel.temperature = 1200.0 # temperature in Kelvin
Warning
MCNP users should be aware that OpenMC does not use the concept of
cross section suffixes like “71c” or “80c”. Temperatures in Kelvin
should be assigned directly per material or per cell using the
Material.temperature
or Cell.temperature
attributes, respectively.
5.6. Material Mixtures¶
In OpenMC it is possible to mix any number of materials to create a new material
with the correct nuclide composition and density. The
Material.mix_materials()
method takes a list of materials and
a list of their mixing fractions. Mixing fractions can be provided as atomic
fractions, weight fractions, or volume fractions. The fraction type
can be specified by passing ‘ao’, ‘wo’, or ‘vo’ as the third argument, respectively.
For example, assuming the required materials have already been defined, a MOX
material with 3% plutonium oxide by weight could be created using the following:
mox = openmc.Material.mix_materials([uo2, puo2], [0.97, 0.03], 'wo')
It should be noted that, if mixing fractions are specifed as atomic or weight fractions, the supplied fractions should sum to one. If the fractions are specified as volume fractions, and the sum of the fractions is less than one, then the remaining fraction is set as void material.
Warning
Materials with \(S(\alpha,\beta)\) thermal scattering data
cannot be used in Material.mix_materials()
. However, thermal
scattering data can be added to a material created by
Material.mix_materials()
.
5.7. Material Collections¶
The openmc executable expects to find a materials.xml
file
when it is run. To create this file, one needs to instantiate the
openmc.Materials
class and add materials to it. The Materials
class acts like a list (in fact, it is a subclass of Python’s built-in
list
class), so materials can be added by passing a list to the
constructor, using methods like append()
, or through the operator
+=
. Once materials have been added to the collection, it can be exported
using the Materials.export_to_xml()
method.
materials = openmc.Materials()
materials.append(water)
materials += [uo2, zircaloy]
materials.export_to_xml()
# This is equivalent
materials = openmc.Materials([water, uo2, zircaloy])
materials.export_to_xml()
5.7.1. Cross Sections¶
OpenMC uses a file called cross_sections.xml to
indicate where cross section data can be found on the filesystem. This file
serves the same role that xsdir
does for MCNP or xsdata
does for
Serpent. Information on how to generate a cross section listing file can be
found in Manually Creating a Library from ACE files. Once you have a cross sections file that has
been generated, you can tell OpenMC to use this file either by setting
Materials.cross_sections
or by setting the
OPENMC_CROSS_SECTIONS
environment variable to the path of the
cross_sections.xml
file. The former approach would look like:
materials.cross_sections = '/path/to/cross_sections.xml'