.. _usersguide_scripts: ======================= Executables and Scripts ======================= .. _scripts_openmc: ---------- ``openmc`` ---------- Once you have a model built (see :ref:`usersguide_basics`), you can either run the openmc executable directly from the directory containing your XML input files, or you can specify as a command-line argument the directory containing the XML input files. For example, if your XML input files are in the directory ``/home/username/somemodel/``, one way to run the simulation would be: .. code-block:: sh cd /home/username/somemodel openmc Alternatively, you could run from any directory: .. code-block:: sh openmc /home/username/somemodel Note that in the latter case, any output files will be placed in the present working directory which may be different from ``/home/username/somemodel``. ``openmc`` accepts the following command line flags: -c, --volume Run in stochastic volume calculation mode -g, --geometry-debug Run in geometry debugging mode, where cell overlaps are checked for after each move of a particle -n, --particles N Use *N* particles per generation or batch -p, --plot Run in plotting mode -r, --restart file Restart a previous run from a state point or a particle restart file -s, --threads N Run with *N* OpenMP threads -t, --track Write tracks for all particles -v, --version Show version information -h, --help Show help message .. note:: If you're using the Python API, :func:`openmc.run` is equivalent to running ``openmc`` from the command line. .. _scripts_ace: ---------------------- ``openmc-ace-to-hdf5`` ---------------------- This script can be used to create HDF5 nuclear data libraries used by OpenMC if you have existing ACE files. There are four different ways you can specify ACE libraries that are to be converted: 1. List each ACE library as a positional argument. This is very useful in conjunction with the usual shell utilities (``ls``, ``find``, etc.). 2. Use the ``--xml`` option to specify a pre-v0.9 cross_sections.xml file. 3. Use the ``--xsdir`` option to specify a MCNP xsdir file. 4. Use the ``--xsdata`` option to specify a Serpent xsdata file. The script does not use any extra information from cross_sections.xml/ xsdir/ xsdata files to determine whether the nuclide is metastable. Instead, the ``--metastable`` argument can be used to specify whether the ZAID naming convention follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data convention (essentially the same as NNDC, except that the first metastable state of Am242 is 95242 and the ground state is 95642). The optional ``--fission_energy_release`` argument will accept an HDF5 file containing a library of fission energy release (ENDF MF=1 MT=458) data. A library built from ENDF/B-VII.1 data is released with OpenMC and can be found at openmc/data/fission_Q_data_endb71.h5. This data is necessary for 'fission-q-prompt' and 'fission-q-recoverable' tallies, but is not needed otherwise. -h, --help show help message and exit -d DESTINATION, --destination DESTINATION Directory to create new library in -m META, --metastable META How to interpret ZAIDs for metastable nuclides. META can be either 'nndc' or 'mcnp'. (default: nndc) --xml XML Old-style cross_sections.xml that lists ACE libraries --xsdir XSDIR MCNP xsdir file that lists ACE libraries --xsdata XSDATA Serpent xsdata file that lists ACE libraries --fission_energy_release FISSION_ENERGY_RELEASE HDF5 file containing fission energy release data .. _scripts_mcnp70: ------------------------------ ``openmc-convert-mcnp70-data`` ------------------------------ This script converts ENDF/B-VII.0 ACE data from the MCNP5/6 distribution into an HDF5 library that can be used by OpenMC. This assumes that you have a directory containing files named endf70a, endf70b, ..., endf70k, and endf70sab. The path to the directory containing these files should be given as a positional argument. The following optional arguments are available: -d DESTINATION, --destination DESTINATION Directory to create new library in (Default: mcnp_endfb70) .. _scripts_mcnp71: ------------------------------ ``openmc-convert-mcnp71-data`` ------------------------------ This script converts ENDF/B-VII.1 ACE data from the MCNP6 distribution into an HDF5 library that can be used by OpenMC. This assumes that you have a directory containing subdirectories 'endf71x' and 'ENDF71SaB'. The path to the directory containing these subdirectories should be given as a positional argument. The following optional arguments are available: -d DESTINATION, --destination DESTINATION Directory to create new library in (Default: mcnp_endfb71) -f FER, --fission_energy_release FER HDF5 file containing fission energy release data .. _scripts_jeff: ------------------------ ``openmc-get-jeff-data`` ------------------------ This script downloads `JEFF 3.2 ACE data `_ from OECD/NEA and converts it to a multi-temperature HDF5 library for use with OpenMC. It has the following optional arguments: -b, --batch Suppress standard in -d DESTINATION, --destination DESTINATION Directory to create new library in (default: jeff-3.2-hdf5) .. warning:: This script will download approximately 9 GB of data. Extracting and processing the data may require as much as 40 GB of additional free disk space. .. _scripts_multipole: ----------------------------- ``openmc-get-multipole-data`` ----------------------------- This script downloads and extracts windowed multipole data based on ENDF/B-VII.1. It has the following optional arguments: -b, --batch Suppress standard in .. _scripts_nndc: ------------------------ ``openmc-get-nndc-data`` ------------------------ This script downloads `ENDF/B-VII.1 ACE data `_ from NNDC and converts it to an HDF5 library for use with OpenMC. This script has the following optional arguments: -b, --batch Suppress standard in ----------------------- ``openmc-make-compton`` ----------------------- This script generates an HDF5 file called ``compton_profiles.h5`` that contains Compton profile data using an existing data library from `Geant4 `_. Note that OpenMC includes this data file by default so it should not be necessary in practice to generate it yourself. .. _scripts_plot: -------------------------- ``openmc-plot-mesh-tally`` -------------------------- ``openmc-plot-mesh-tally`` provides a graphical user interface for plotting mesh tallies. The path to the statepoint file can be provided as an optional arugment (if omitted, a file dialog will be presented). .. _scripts_track: ----------------------- ``openmc-track-to-vtk`` ----------------------- This script converts HDF5 :ref:`particle track files ` to VTK poly data that can be viewed with ParaView or VisIt. The filenames of the particle track files should be given as posititional arguments. The output filename can also be changed with the ``-o`` flag: -o OUT, --out OUT Output VTK poly filename ------------------------ ``openmc-update-inputs`` ------------------------ If you have existing XML files that worked in a previous version of OpenMC that no longer work with the current version, you can try to update these files using ``openmc-update-inputs``. If any of the given files do not match the most up-to-date formatting, then they will be automatically rewritten. The old out-of-date files will not be deleted; they will be moved to a new file with '.original' appended to their name. Formatting changes that will be made: geometry.xml Lattices containing 'outside' attributes/tags will be replaced with lattices containing 'outer' attributes, and the appropriate cells/universes will be added. Any 'surfaces' attributes/elements on a cell will be renamed 'region'. materials.xml Nuclide names will be changed from ACE aliases (e.g., Am-242m) to HDF5/GND names (e.g., Am242_m1). Thermal scattering table names will be changed from ACE aliases (e.g., HH2O) to HDF5/GND names (e.g., c_H_in_H2O). ---------------------- ``openmc-update-mgxs`` ---------------------- This script updates OpenMC's deprecated multi-group cross section XML files to the latest HDF5-based format. -i IN, --input IN Input XML file -o OUT, --output OUT Output file in HDF5 format .. _scripts_validate: ----------------------- ``openmc-validate-xml`` ----------------------- Input files can be checked before executing OpenMC using the ``openmc-validate-xml`` script which is installed alongside the Python API. Two command line arguments can be set when running ``openmc-validate-xml``: -i, --input-path Location of OpenMC input files. -r, --relaxng-path Location of OpenMC RelaxNG files If the RelaxNG path is not set, the script will search for these files because it expects that the user is either running the script located in the install directory ``bin`` folder or in ``src/utils``. Once executed, it will match OpenMC XML files with their RelaxNG schema and check if they are valid. Below is a table of the messages that will be printed after each file is checked. ======================== =================================== Message Description ======================== =================================== [XML ERROR] Cannot parse XML file. [NO RELAXNG FOUND] No RelaxNG file found for XML file. [NOT VALID] XML file does not match RelaxNG. [VALID] XML file matches RelaxNG. ======================== =================================== .. _scripts_voxel: --------------------------- ``openmc-voxel-to-silovtk`` --------------------------- When OpenMC generates :ref:`voxel plots `, they are in an :ref:`HDF5 format ` that is not terribly useful by itself. The ``openmc-voxel-to-silovtk`` script converts a voxel HDF5 file to `VTK `_ or `SILO `_ file. For VTK, you need to have the VTK Python bindings installed. For SILO, you need to have `silomesh `_ installed. To convert a voxel file, simply provide the path to the file: .. code-block:: sh openmc-voxel-to-silovtk voxel_1.h5 The ``openmc-voxel-to-silovtk`` script also takes the following optional command-line arguments: -o, --output Path to output VTK or SILO file -s, --silo Flag to convert to SILO instead of VTK