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177 lines
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ReStructuredText
177 lines
7.2 KiB
ReStructuredText
.. _usersguide_materials:
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.. currentmodule:: openmc
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=====================
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Material Compositions
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=====================
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Materials in OpenMC are defined as a set of nuclides/elements at specified
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densities and are created using the :class:`openmc.Material` class. Once a
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material has been instantiated, nuclides can be added with
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:meth:`Material.add_nuclide` and elements can be added with
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:meth:`Material.add_element`. Densities can be specified using atom fractions or
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weight fractions. For example, to create a material and add Gd152 at 0.5 atom
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percent, you'd run::
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mat = openmc.Material()
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mat.add_nuclide('Gd152', 0.5, 'ao')
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The third argument to :meth:`Material.add_nuclide` can also be 'wo' for weight
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percent. The densities specified for each nuclide/element are relative and are
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renormalized based on the total density of the material. The total density is
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set using the :meth:`Material.set_density` method. The density can be specified
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in gram per cubic centimeter ('g/cm3'), atom per barn-cm ('atom/b-cm'), or
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kilogram per cubic meter ('kg/m3'), e.g.,
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::
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mat.set_density('g/cm3', 4.5)
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----------------
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Natural Elements
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----------------
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The :meth:`Material.add_element` method works exactly the same as
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:meth:`Material.add_nuclide`, except that instead of specifying a single isotope
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of an element, you specify the element itself. For example,
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::
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mat.add_element('C', 1.0)
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Internally, OpenMC stores data on the atomic masses and natural abundances of
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all known isotopes and then uses this data to determine what isotopes should be
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added to the material. When the material is later exported to XML for use by the
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:ref:`scripts_openmc` executable, you'll see that any natural elements were
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expanded to the naturally-occurring isotopes.
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The :meth:`Material.add_element` method can also be used to add uranium at a
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specified enrichment through the `enrichment` argument. For example, the
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following would add 3.2% enriched uranium to a material::
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mat.add_element('U', 1.0, enrichment=3.2)
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In addition to U235 and U238, concentrations of U234 and U236 will be present
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and are determined through a correlation based on measured data.
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Often, cross section libraries don't actually have all naturally-occurring
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isotopes for a given element. For example, in ENDF/B-VII.1, cross section
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evaluations are given for O16 and O17 but not for O18. If OpenMC is aware of
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what cross sections you will be using (through the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable), it will attempt to only
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put isotopes in your model for which you have cross section data. In the case of
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oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped with O16.
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-----------------------
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Thermal Scattering Data
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-----------------------
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If you have a moderating material in your model like water or graphite, you
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should assign thermal scattering data (so-called :math:`S(\alpha,\beta)`) using
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the :meth:`Material.add_s_alpha_beta` method. For example, to model light water,
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you would need to add hydrogen and oxygen to a material and then assign the
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``c_H_in_H2O`` thermal scattering data::
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water = openmc.Material()
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water.add_nuclide('H1', 2.0)
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water.add_nuclide('O16', 1.0)
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water.add_s_alpha_beta('c_H_in_H2O')
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water.set_density('g/cm3', 1.0)
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.. _usersguide_naming:
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------------------
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Naming Conventions
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------------------
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OpenMC uses the GND_ naming convention for nuclides, metastable states, and
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compounds:
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:Nuclides: ``SymA`` where "A" is the mass number (e.g., ``Fe56``)
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:Elements: ``Sym0`` (e.g., ``Fe0`` or ``C0``)
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:Excited states: ``SymA_eN`` (e.g., ``V51_e1`` for the first excited state of
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Vanadium-51.) This is only used in decay data.
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:Metastable states: ``SymA_mN`` (e.g., ``Am242_m1`` for the first excited state
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of Americium-242).
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:Compounds: ``c_String_Describing_Material`` (e.g., ``c_H_in_H2O``). Used for
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thermal scattering data.
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.. important:: The element syntax, e.g., ``C0``, is only used when the cross
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section evaluation is an elemental evaluation, like carbon in
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ENDF/B-VII.1! If you are adding an element via
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:meth:`Material.add_element`, just use ``Sym``.
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.. _GND: https://www.oecd-nea.org/science/wpec/sg38/Meetings/2016_May/tlh4gnd-main.pdf
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-----------
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Temperature
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-----------
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Some Monte Carlo codes define temperature implicitly through the cross section
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data, which is itself given only at a particular temperature. In OpenMC, the
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material definition is decoupled from the specification of temperature. Instead,
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temperatures are assigned to :ref:`cells <usersguide_cells>`
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directly. Alternatively, a default temperature can be assigned to a material
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that is to be applied to any cell where the material is used. In the absence of
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any cell or material temperature specification, a global default temperature can
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be set that is applied to all cells and materials. Anytime a material
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temperature is specified, it will override the global default
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temperature. Similarly, anytime a cell temperatures is specified, it will
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override the material or global default temperature. All temperatures should be
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given in units of Kelvin.
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To assign a default material temperature, one should use the ``temperature``
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attribute, e.g.,
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::
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hot_fuel = openmc.Material()
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hot_fuel.temperature = 1200.0 # temperature in Kelvin
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.. warning:: MCNP_ users should be aware that OpenMC does not use the concept of
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cross section suffixes like "71c" or "80c". Temperatures in Kelvin
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should be assigned directly per material or per cell using the
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:attr:`Material.temperature` or :attr:`Cell.temperature`
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attributes, respectively.
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--------------------
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Material Collections
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--------------------
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The :ref:`scripts_openmc` executable expects to find a ``materials.xml`` file
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when it is run. To create this file, one needs to instantiate the
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:class:`openmc.Materials` class and add materials to it. The :class:`Materials`
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class acts like a list (in fact, it is a subclass of Python's built-in
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:class:`list` class), so materials can be added by passing a list to the
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constructor, using methods like ``append()``, or through the operator
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``+=``. Once materials have been added to the collection, it can be exported
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using the :meth:`Materials.export_to_xml` method.
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::
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materials = openmc.Materials()
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materials.append(water)
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materials += [uo2, zircaloy]
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materials.export_to_xml()
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# This is equivalent
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materials = openmc.Materials([water, uo2, zircaloy])
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materials.export_to_xml()
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Cross Sections
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--------------
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OpenMC uses a file called :ref:`cross_sections.xml <io_cross_sections>` to
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indicate where cross section data can be found on the filesystem. This file
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serves the same role that ``xsdir`` does for MCNP_ or ``xsdata`` does for
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Serpent. Information on how to generate a cross section listing file can be
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found in :ref:`create_xs_library`. Once you have a cross sections file that has
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been generated, you can tell OpenMC to use this file either by setting
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:attr:`Materials.cross_sections` or by setting the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the path of the
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``cross_sections.xml`` file. The former approach would look like::
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materials.cross_sections = '/path/to/cross_sections.xml'
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.. _MCNP: https://mcnp.lanl.gov/
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