Nuclear data processing capabilities in OpenMC
1 Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439, USA
2 Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139, USA
a e-mail: firstname.lastname@example.org
Published online: 13 September 2017
This work describes newly developed features of the OpenMC code for nuclear data processing. OpenMC, in addition to being a transport code, includes a rich, extensible Python API that enables programmatic pre- and post-processing. A new openmc.data package in the Python API enables users to parse ENDF and ACE files and convert them to an HDF5 format. With this capability, the OpenMC transport solver now relies on HDF5 nuclear data files rather than ACE files produced from NJOY. Moving to a native format will give much greater flexibility for researchers to explore new methods and algorithms that rely on storing data that is not present in the ACE format. Additionally, the module may serve as an independent implementation of the proposed Generalized Nuclear Data (GND) format in the future.
© The Authors, published by EDP Sciences, 2017
This is an Open Access article distributed under the terms of the Creative Commons Attribution License 4.0, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.