# 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. 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
openmc


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 -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, 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 -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

## 11.3. 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.

## 11.4. openmc-make-depletion-chain¶

This script generates a depletion chain file called chain_endfb71.xml using ENDF/B-VII.1 nuclear data. If the OPENMC_ENDF_DATA variable is not set, and "neutron", "decay", "nfy" directories do not exist, then ENDF/B-VII.1 data will be downloaded.

## 11.5. openmc-make-depletion-chain-casl¶

This script generates a depletion chain called chain_casl.xml using ENDF/B-VII.1 nuclear data for a simplified chain. The nuclides were chosen by CASL-ORIGEN, which can be found in Appendix A of Kang Seog Kim, “Specification for the VERA Depletion Benchmark Suite”, CASL-U-2015-1014-000, Rev. 0, ORNL/TM-2016/53, 2016. Te129 has been added into this chain due to its link to I129 production.

If the OPENMC_ENDF_DATA variable is not set, and "neutron", "decay", "nfy" directories to not exist, then ENDF/B-VII.1 data will be downloaded.

## 11.6. openmc-make-stopping-powers¶

This script generates an HDF5 file called stopping_power.h5 that contains radiative and collision stopping powers and mean excitation energy pulled from the NIST ESTAR database. Note that OpenMC includes this data file by default so it should not be necessary in practice to generate it yourself.

## 11.7. 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).

## 11.8. 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

## 11.9. 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).

## 11.10. 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

## 11.11. 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.

## 11.12. 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