11. Executables and Scripts

11.1. openmc

Once you have a model built (see Basics of Using OpenMC), 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.


OpenMC models should be treated as code, and it is important to be careful with code from untrusted sources.

For example, if your XML input files are in the directory /home/username/somemodel/, one way to run the simulation would be:

cd /home/username/somemodel

Alternatively, you could run from any directory:

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

-e, --event

Run using event-based parallelism

-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 (up to max_tracks)

-v, --version

Show version information

-h, --help

Show help message


If you’re using the Python API, openmc.run() is equivalent to running openmc from the command line.

11.2. 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


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

11.3. 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).

11.4. openmc-track-combine

This script combines multiple HDF5 particle track files into a single HDF5 particle track file. 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 HDF5 particle track file

11.5. openmc-track-to-vtk

This script converts HDF5 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

11.6. 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:


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’.


Nuclide names will be changed from ACE aliases (e.g., Am-242m) to HDF5/GNDS names (e.g., Am242_m1). Thermal scattering table names will be changed from ACE aliases (e.g., HH2O) to HDF5/GNDS names (e.g., c_H_in_H2O).

11.7. 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

11.8. openmc-voxel-to-vtk

When OpenMC generates voxel plots, they are in an HDF5 format that is not terribly useful by itself. The openmc-voxel-to-vtk script converts a voxel HDF5 file to a VTK file. To run this script, you will need to have the VTK Python bindings installed. To convert a voxel file, simply provide the path to the file:

openmc-voxel-to-vtk voxel_1.h5

The openmc-voxel-to-vtk script also takes the following optional command-line arguments:

-o, --output

Path to output VTK file