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.. _usersguide_cross_sections:
===========================
Cross Section Configuration
===========================
In order to run a simulation with OpenMC, you will need cross section data for
each nuclide or material in your problem. OpenMC can be run in continuous-energy
or multi-group mode.
In continuous-energy mode, OpenMC uses a native `HDF5
<https://support.hdfgroup.org/HDF5/>`_ format (see :ref:`io_nuclear_data`) to
store all nuclear data. Pregenerated HDF5 libraries can be found at
https://openmc.org; unless you have specific data needs, it is highly
recommended to use one of the pregenerated libraries. Alternatively, if you have
ACE format data that was produced with NJOY_, such as that distributed with
MCNP_ or Serpent_, it can be converted to the HDF5 format using the :ref:`using
the Python API <create_xs_library>`. Several sources provide openly available
ACE data including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the
`LANL Nuclear Data Team <https://nucleardata.lanl.gov/>`_. In addition to
tabulated cross sections in the HDF5 files, OpenMC relies on :ref:`windowed
multipole <windowed_multipole>` data to perform on-the-fly Doppler broadening.
In multi-group mode, OpenMC utilizes an HDF5-based library format which can be
used to describe nuclide- or material-specific quantities.
---------------------
Environment Variables
---------------------
When :ref:`scripts_openmc` is run, it will look for several environment
variables that indicate where cross sections can be found. While the location of
cross sections can also be indicated through the
:attr:`openmc.Materials.cross_setion` attribute (or in the :ref:`materials.xml
<io_materials>` file), if you always use the same set of cross section data, it
is often easier to just set an environment variable that will be picked up by
default every time OpenMC is run. The following environment variables are used:
:envvar:`OPENMC_CROSS_SECTIONS`
Indicates the path to the :ref:`cross_sections.xml <io_cross_sections>`
summary file that is used to locate HDF5 format cross section libraries if the
user has not specified :attr:`Materials.cross_sections` (equivalently, the
:ref:`cross_sections` in :ref:`materials.xml <io_materials>`).
:envvar:`OPENMC_MG_CROSS_SECTIONS`
Indicates the path to the an :ref:`HDF5 file <io_mgxs_library>` that contains
multi-group cross sections if the user has not specified
:attr:`Materials.cross_sections` (equivalently, the :ref:`cross_sections` in
:ref:`materials.xml <io_materials>`).
To set these environment variables persistently, export them from your shell
profile (``.profile`` or ``.bashrc`` in bash_).
.. _bash: http://www.linuxfromscratch.org/blfs/view/6.3/postlfs/profile.html
--------------------------------
Continuous-Energy Cross Sections
--------------------------------
Using Pregenerated Libraries
----------------------------
Various evaluated nuclear data libraries have been processed into the HDF5
format required by OpenMC and can be found at https://openmc.org. You
can find both libraries generated by the OpenMC development team as well as
libraries based on ACE files distributed elsewhere. To use these libraries,
download the archive file, unpack it, and then set your
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
the ``cross_sections.xml`` file contained in the unpacked directory.
.. _create_xs_library:
Manually Creating a Library from ACE files
------------------------------------------
.. currentmodule:: openmc.data
The scripts described above use the :mod:`openmc.data` module in the Python API
to convert ACE data and create a :ref:`cross_sections.xml <io_cross_sections>`
file. For those who prefer to use the API directly, the
:class:`openmc.data.IncidentNeutron` and :class:`openmc.data.ThermalScattering`
classes can be used to read ACE data and convert it to HDF5. For
continuous-energy incident neutron data, use the
:meth:`IncidentNeutron.from_ace` class method to read in an existing ACE file
and the :meth:`IncidentNeutron.export_to_hdf5` method to write the data to an
HDF5 file.
::
u235 = openmc.data.IncidentNeutron.from_ace('92235.710nc')
u235.export_to_hdf5('U235.h5')
If you have multiple ACE files for the same nuclide at different temperatures,
you can use the :meth:`IncidentNeutron.add_temperature_from_ace` method to
append cross sections to an existing :class:`IncidentNeutron` instance::
u235 = openmc.data.IncidentNeutron.from_ace('92235.710nc')
for suffix in [711, 712, 713, 714, 715, 716]:
u235.add_temperature_from_ace('92235.{}nc'.format(suffix))
u235.export_to_hdf5('U235.h5')
Similar methods exist for thermal scattering data:
::
light_water = openmc.data.ThermalScattering.from_ace('lwtr.20t')
for suffix in range(21, 28):
light_water.add_temperature_from_ace('lwtr.{}t'.format(suffix))
light_water.export_to_hdf5('lwtr.h5')
Once you have created corresponding HDF5 files for each of your ACE files, you
can create a library and export it to XML using the
:class:`openmc.data.DataLibrary` class::
library = openmc.data.DataLibrary()
library.register_file('U235.h5')
library.register_file('lwtr.h5')
...
library.export_to_xml()
At this point, you will have a ``cross_sections.xml`` file that you can use in
OpenMC.
.. hint:: The :class:`IncidentNeutron` class allows you to view/modify cross
sections, secondary angle/energy distributions, probability tables,
etc. For a more thorough overview of the capabilities of this class,
see the :ref:`notebook_nuclear_data` example notebook.
Manually Creating a Library from ENDF files
-------------------------------------------
If you need to create a nuclear data library and you do not already have
suitable ACE files or you need to further customize the data (for example,
adding more temperatures), the :meth:`IncidentNeutron.from_njoy` and
:meth:`ThermalScattering.from_njoy` methods can be used to create data instances
by directly running NJOY_. Both methods require that you pass the name of ENDF
file(s) that are passed on to NJOY. For example, to generate data for Zr-92::
zr92 = openmc.data.IncidentNeutron.from_njoy('n-040_Zr_092.endf')
By default, data is produced at room temperature, 293.6 K. You can also specify
a list of temperatures that you want data at::
zr92 = openmc.data.IncidentNeutron.from_njoy(
'n-040_Zr_092.endf', temperatures=[300., 600., 1000.])
The :meth:`IncidentNeutron.from_njoy` method assumes you have an executable
named ``njoy`` available on your path. If you want to explicitly name the
executable, the ``njoy_exec`` optional argument can be used. Additionally, the
``stdout`` argument can be used to show the progress of the NJOY run.
To generate a thermal scattering file, you need to specify both an ENDF incident
neutron sub-library file as well as a thermal neutron scattering sub-library
file; for example::
light_water = openmc.data.ThermalScattering.from_njoy(
'neutrons/n-001_H_001.endf', 'thermal_scatt/tsl-HinH2O.endf')
Once you have instances of :class:`IncidentNeutron` and
:class:`ThermalScattering`, a library can be created by using the
``export_to_hdf5()`` methods and the :class:`DataLibrary` class as described in
:ref:`create_xs_library`.
Enabling Resonance Scattering Treatments
----------------------------------------
In order for OpenMC to correctly treat elastic scattering in heavy nuclides
where low-lying resonances might be present (see
:ref:`energy_dependent_xs_model`), the elastic scattering cross section at 0 K
must be present. If the data you are using was generated via
:meth:`IncidentNeutron.from_njoy`, you will already have 0 K elastic scattering
cross sections available. Otherwise, to add 0 K elastic scattering cross
sections to an existing :class:`IncidentNeutron` instance, you can use the
:meth:`IncidentNeutron.add_elastic_0K_from_endf` method which requires an ENDF
file for the nuclide you are modifying::
u238 = openmc.data.IncidentNeutron.from_hdf5('U238.h5')
u238.add_elastic_0K_from_endf('n-092_U_238.endf')
u238.export_to_hdf5('U238_with_0K.h5')
With 0 K elastic scattering data present, you can turn on a resonance scattering
method using :attr:`Settings.resonance_scattering`.
.. note:: The process of reconstructing resonances and generating tabulated 0 K
cross sections can be computationally expensive, especially for
nuclides like U-238 where thousands of resonances are present. Thus,
running the :meth:`IncidentNeutron.add_elastic_0K_from_endf` method
may take several minutes to complete.
Photon Cross Sections
---------------------
Photon interaction data is needed to run OpenMC with photon transport enabled.
Some of this data, namely bremsstrahlung cross sections from `Seltzer and
Berger`_, mean excitation energy from the `NIST ESTAR database`_, and Compton
profiles calculated by `Biggs et al.`_ and available in the Geant4 G4EMLOW data
file, is distributed with OpenMC. The rest is available from the NNDC_, which
provides ENDF data from the photo-atomic and atomic relaxation sublibraries of
the ENDF/B-VII.1 library.
Most of the pregenerated HDF5 libraries available at https://openmc.org
already have photon interaction data included. If you are building a data
library yourself, it is possible to use the Python API directly to convert
photon interaction data from an ENDF or ACE file to an HDF5 file. The
:class:`openmc.data.IncidentPhoton` class contains an
:meth:`IncidentPhoton.from_ace` method that will generate photon data from an
ACE table and an :meth:`IncidentPhoton.export_to_hdf5` method that writes the
data to an HDF5 file:
::
u = openmc.data.IncidentPhoton.from_ace('92000.12p')
u.export_to_hdf5('U.h5')
Similarly, the :meth:`IncidentPhoton.from_endf` method can be used to read
photon data from an ENDF file. In this case, both the photo-atomic and atomic
relaxation sublibrary files are required:
::
u = openmc.data.IncidentPhoton.from_endf('photoat-092_U_000.endf',
'atom-092_U_000.endf')
Once the HDF5 files have been generated, a library can be created using the
:class:`DataLibrary` class as described in :ref:`create_xs_library`.
-----------------------
Windowed Multipole Data
-----------------------
OpenMC is capable of using windowed multipole data for on-the-fly Doppler
broadening. A comprehensive multipole data library containing all nuclides in
ENDF/B-VII.1 is available on `GitHub
<https://github.com/mit-crpg/WMP_Library>`_. To obtain this library, download
and unpack an archive (.zip or .tag.gz) from GitHub. Once unpacked, you can use
the :class:`openmc.data.DataLibrary` class to register the .h5 files as
described in :ref:`create_xs_library`.
The `official ENDF/B-VII.1 HDF5 library
<https://openmc.org/official-data-libraries/>`_ includes the windowed
multipole library, so if you are using this library, the windowed multipole data
will already be available to you.
--------------------------
Multi-Group Cross Sections
--------------------------
Multi-group cross section libraries are generally tailored to the specific
calculation to be performed. Therefore, at this point in time, OpenMC is not
distributed with any pre-existing multi-group cross section libraries.
However, if obtained or generated their own library, the user
should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
to the absolute path of the file library expected to used most frequently.
For an example of how to create a multi-group library, see
:ref:`notebook_mg_mode_part_i`.
.. _NJOY: http://www.njoy21.io/
.. _NNDC: https://www.nndc.bnl.gov/endf
.. _MCNP: https://mcnp.lanl.gov
.. _Serpent: http://montecarlo.vtt.fi
.. _ENDF/B: https://www.nndc.bnl.gov/endf/b7.1/acefiles.html
.. _JEFF: http://www.oecd-nea.org/dbdata/jeff/jeff33/
.. _TENDL: https://tendl.web.psi.ch/tendl_2017/tendl2017.html
.. _Seltzer and Berger: https://doi.org/10.1016/0092-640X(86)90014-8
.. _NIST ESTAR database: https://physics.nist.gov/PhysRefData/Star/Text/ESTAR.html
.. _Biggs et al.: https://doi.org/10.1016/0092-640X(75)90030-3