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Merge pull request #2211 from paulromano/config
Allow configuration of data sources via openmc.config
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Jonathan Shimwell
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16 changed files with 299 additions and 172 deletions
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@ -1,55 +1,73 @@
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.. _usersguide_cross_sections:
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.. _usersguide_data:
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===========================
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Cross Section Configuration
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===========================
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==================
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Data Configuration
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==================
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In order to run a simulation with OpenMC, you will need cross section data for
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each nuclide or material in your problem. OpenMC can be run in continuous-energy
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or multi-group mode.
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OpenMC relies on a variety of physical data in order to carry out transport
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simulations, depletion simulations, and other common tasks. As a user, you are
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responsible for specifying one or more of the following:
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In continuous-energy mode, OpenMC uses a native `HDF5
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<https://support.hdfgroup.org/HDF5/>`_ format (see :ref:`io_nuclear_data`) to
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store all nuclear data. Pregenerated HDF5 libraries can be found at
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https://openmc.org; unless you have specific data needs, it is highly
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recommended to use one of the pregenerated libraries. Alternatively, if you have
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ACE format data that was produced with NJOY_, such as that distributed with
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MCNP_ or Serpent_, it can be converted to the HDF5 format using the :ref:`using
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the Python API <create_xs_library>`. Several sources provide openly available
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ACE data including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the
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`LANL Nuclear Data Team <https://nucleardata.lanl.gov/>`_. In addition to
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tabulated cross sections in the HDF5 files, OpenMC relies on :ref:`windowed
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multipole <windowed_multipole>` data to perform on-the-fly Doppler broadening.
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- **Cross sections (XML)** -- A :ref:`cross sections XML <io_cross_sections>`
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file (commonly named ``cross_sections.xml``) contains a listing of other data
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files, in particular neutron cross sections, photon cross sections, and
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windowed multipole data. Each of those files, in turn, uses a `HDF5
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<https://support.hdfgroup.org/HDF5/>`_ format (see :ref:`io_nuclear_data`). In
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order to run transport simulations with continuous-energy cross sections, you
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need to specify this file.
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In multi-group mode, OpenMC utilizes an HDF5-based library format which can be
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used to describe nuclide- or material-specific quantities.
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- **Depletion chain (XML)** -- A :ref:`depletion chain XML <io_depletion_chain>`
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file contains decay data, fission product yields, and information on what
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neutron reactions can result in transmutation. This file is needed for
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depletion/activation calculations as well as some basic functions in the
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:mod:`openmc.data` module.
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- **Multigroup cross sections (HDF5)** -- OpenMC can also perform transport
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simulations using multigroup data. In this case, multigroup cross sections are
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stored in a single :ref:`HDF5 file <io_mgxs_library>`. Thus, in order to run a
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multigroup transport simulation, this file needs to be specified.
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Each of the above files can specified in several ways. In the Python API, a
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:ref:`runtime configuration variable <usersguide_data_runtime>`
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:data:`openmc.config` can be used to specify any of the above and is initialized
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using a set of environment variables.
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.. _usersguide_data_runtime:
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---------------------
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Environment Variables
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Runtime Configuration
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---------------------
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When :ref:`scripts_openmc` is run, it will look for several environment
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variables that indicate where cross sections can be found. While the location of
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cross sections can also be indicated through the
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:attr:`openmc.Materials.cross_sections` attribute (or in the :ref:`materials.xml
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<io_materials>` file), if you always use the same set of cross section data, it
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is often easier to just set an environment variable that will be picked up by
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default every time OpenMC is run. The following environment variables are used:
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Data sources for OpenMC can be specified at runtime in Python using the
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:data:`openmc.config` variable. This variable acts like a dictionary and stores
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key-values pairs, where the values are file paths (strings or path-like objects)
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and the key can be one of the following:
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:envvar:`OPENMC_CROSS_SECTIONS`
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Indicates the path to the :ref:`cross_sections.xml <io_cross_sections>`
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summary file that is used to locate HDF5 format cross section libraries if the
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user has not specified :attr:`openmc.Materials.cross_sections` (equivalently,
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the :ref:`cross_sections` in :ref:`materials.xml <io_materials>`).
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``"cross_sections"``
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Indicates the path to the :ref:`cross sections XML <io_cross_sections>` file
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that lists HDF5 format neutron cross sections, photon cross sections, and
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windowed multipole data. At startup, this is initialized with the value of the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable. Note that the
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:attr:`openmc.Materials.cross_sections` attribute will override this, if
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specified.
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:envvar:`OPENMC_MG_CROSS_SECTIONS`
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``"chain_file"``
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Indicates the path to the :ref:`depletion chain XML <io_depletion_chain>` file
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that contains decay data, fission product yields, and what neutron reactions
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may result in transmutation of a target nuclide. At startup, this is
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initialized with the value of the :envvar:`OPENMC_CHAIN_FILE` environment
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variable.
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``"mg_cross_sections"``
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Indicates the path to an :ref:`HDF5 file <io_mgxs_library>` that contains
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multi-group cross sections if the user has not specified
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:attr:`openmc.Materials.cross_sections` (equivalently, the
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:ref:`cross_sections` in :ref:`materials.xml <io_materials>`).
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multigroup cross sections. At startup, this is initialized with the value of
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the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable. Note that the
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:attr:`openmc.Materials.cross_sections` attribute will override this if
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specified.
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To set these environment variables persistently, export them from your shell
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profile (``.profile`` or ``.bashrc`` in bash_).
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If you want to persistently set the environment variables used to initialized
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the configuration, export them from your shell profile (``.profile`` or
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``.bashrc`` in bash_).
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.. _bash: http://www.linuxfromscratch.org/blfs/view/6.3/postlfs/profile.html
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@ -61,12 +79,13 @@ Using Pregenerated Libraries
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----------------------------
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Various evaluated nuclear data libraries have been processed into the HDF5
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format required by OpenMC and can be found at https://openmc.org. You
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can find both libraries generated by the OpenMC development team as well as
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libraries based on ACE files distributed elsewhere. To use these libraries,
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download the archive file, unpack it, and then set your
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
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the ``cross_sections.xml`` file contained in the unpacked directory.
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format required by OpenMC and can be found at https://openmc.org. Unless you
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have specific data needs, it is highly recommended to use one of the
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pregenerated libraries. You can find both libraries generated by the OpenMC
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development team as well as libraries based on ACE files distributed elsewhere.
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To use these libraries, download the archive file, unpack it, and then specify
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the path of the ``cross_sections.xml`` file contained in the unpacked directory
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as described in :ref:`usersguide_data_runtime`.
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.. _create_xs_library:
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@ -75,6 +94,12 @@ Manually Creating a Library from ACE files
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.. currentmodule:: openmc.data
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If you have ACE format data that was produced with NJOY_, such as that
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distributed with MCNP_ or Serpent_, it can be converted to the HDF5 format using
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the using the Python API. Several sources provide openly available ACE data
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including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the `LANL
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Nuclear Data Team <https://nucleardata.lanl.gov/>`_.
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The :mod:`openmc.data` module in the Python API enables users to directly
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convert ACE data to OpenMC's HDF5 format and create a corresponding
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:ref:`cross_sections.xml <io_cross_sections>` file. For those who prefer to use
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@ -224,6 +249,19 @@ relaxation sublibrary files are required:
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Once the HDF5 files have been generated, a library can be created using the
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:class:`DataLibrary` class as described in :ref:`create_xs_library`.
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-----------
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Chain Files
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-----------
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Pregenerated depletion chain XML files can be found at https://openmc.org.
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Additionally, depletion chains can be generated using the
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:class:`openmc.deplete.Chain` class. In particular, the
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:meth:`~openmc.deplete.Chain.from_endf` method allows a chain to be generated
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starting from a set of ENDF incident neutron, decay, and fission product yield
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sublibrary files. Once you've downloaded or generated a depletion chain XML
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file, make sure to specify its path as described in
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:ref:`usersguide_data_runtime`.
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-----------------------
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Windowed Multipole Data
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-----------------------
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@ -241,18 +279,16 @@ The `official ENDF/B-VII.1 HDF5 library
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multipole library, so if you are using this library, the windowed multipole data
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will already be available to you.
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--------------------------
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Multi-Group Cross Sections
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--------------------------
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-------------------------
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Multigroup Cross Sections
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-------------------------
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Multi-group cross section libraries are generally tailored to the specific
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Multigroup cross section libraries are generally tailored to the specific
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calculation to be performed. Therefore, at this point in time, OpenMC is not
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distributed with any pre-existing multi-group cross section libraries.
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However, if obtained or generated their own library, the user
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should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
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to the absolute path of the file library expected to used most frequently.
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For an example of how to create a multi-group library, see the `example notebook
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distributed with any pre-existing multigroup cross section libraries. However,
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if a multigroup library file is downloaded or generated, the path to the file
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needs to be specified as described in :ref:`usersguide_data_runtime`. For an
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example of how to create a multigroup library, see the `example notebook
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<../examples/mg-mode-part-i.ipynb>`__.
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.. _NJOY: http://www.njoy21.io/
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@ -13,7 +13,7 @@ essential aspects of using OpenMC to perform simulations.
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beginners
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install
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cross_sections
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data
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basics
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materials
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geometry
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