.. _usersguide_materials:

.. currentmodule:: openmc

=====================
Material Compositions
=====================

Materials in OpenMC are defined as a set of nuclides/elements at specified
densities and are created using the :class:`openmc.Material` class. Once a
material has been instantiated, nuclides can be added with
:meth:`Material.add_nuclide` and elements can be added with
:meth:`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 :meth:`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 :meth:`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)

----------------
Natural Elements
----------------

The :meth:`Material.add_element` method works exactly the same as
:meth:`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)

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
:ref:`scripts_openmc` executable, you'll see that any natural elements are
expanded to the naturally-occurring isotopes.

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 (either through the
:attr:`Materials.cross_sections` attribute or the
:envvar:`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.

-----------------------
Thermal Scattering Data
-----------------------

If you have a moderating material in your model like water or graphite, you
should assign thermal scattering data (so-called :math:`S(\alpha,\beta)`) using
the :meth:`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)

.. _usersguide_naming:

------------------
Naming Conventions
------------------

OpenMC uses the GND_ naming convention for nuclides, metastable states, and
compounds:

:Nuclides: ``SymA`` where "A" is the mass number (e.g., ``Fe56``)
:Elements: ``Sym0`` (e.g., ``Fe0`` or ``C0``)
: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
               :meth:`Material.add_element`, just use ``Sym``.

.. _GND: https://www.oecd-nea.org/science/wpec/sg38/Meetings/2016_May/tlh4gnd-main.pdf

-----------
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 :ref:`cells <usersguide_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
             :attr:`Material.temperature` or :attr:`Cell.temperature`
             attributes, respectively.

--------------------
Material Collections
--------------------

The :ref:`scripts_openmc` executable expects to find a ``materials.xml`` file
when it is run. To create this file, one needs to instantiate the
:class:`openmc.Materials` class and add materials to it. The :class:`Materials`
class acts like a list (in fact, it is a subclass of Python's built-in
:class:`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 :meth:`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()

Cross Sections
--------------

OpenMC uses a file called :ref:`cross_sections.xml <io_cross_sections>` 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 :ref:`create_xs_library`. Once you have a cross sections file that has
been generated, you can tell OpenMC to use this file either by setting
:attr:`Materials.cross_sections` or by setting the
:envvar:`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'

.. _MCNP: https://mcnp.lanl.gov/
