ONIX: an open-source depletion code coupled with OpenMC

Hi All,

I would like to announce the release of ONIX, an open-source depletion code that can be coupled with OpenMC and which has been fully validated.

I started developing ONIX about three years ago with the help of two colleagues, Moritz Kütt and Alexander Glaser. Being a big fan of OpenMC and working on reactors’ fissile material production, I needed a tool to model isotopic evolution in reactors. The development of ONIX became my PhD thesis and we wanted to make sure it would be available to everyone so we decided to make the code open-source.

What can ONIX do?
ONIX can be used in two modes: Standalone and coupled with OpenMC. The standalone mode can simulate the irradiation of material to a neutron flux. The coupled mode allows for full reactor physics simulation (neutron transport + depletion). At the end of a simulation, the output directory contains the evolution of multiple parameters against burnup and days such as k-inf, isotopic densities, power, flux, neutron spectrum, one-group cross sections and isomeric branchings. ONIX has many other functionalities that are described in the documentation (https://onix-documentation.readthedocs.io/en/latest/overview.html)

What nuclear data libraries does ONIX use?
Theoretically, ONIX can be used with any libraries as long as it is converted into a custom format. As of now, ONIX provides ready-to-use ENDF/B-VIII.0 and JEFF3.2 libraries for decay and fission yields. ONIX also provides a constant one-group cross section library obtained from an ONIX+OpenMC coupled simulation of a VVER reactor.

Unlike many other depletion codes, ONIX updates isomeric branching each time OpenMC computes a new neutron spectrum.

What language is ONIX written in?
ONIX is fully written in Python 3 with an object-oriented design. ONIX Python API allows for readability, flexibility and reusability of its functionalities.

What are the applications of ONIX?
ONIX can be used for nuclear reactor simulations, fissile material production analysis, nuclear archaeology and nuclear decay calculations. ONIX is equipped with functionalities that are very useful for nuclear archaeology and nuclear forensic.

How do you install ONIX?
Check out this small installation guide (https://onix-documentation.readthedocs.io/en/latest/installation.html). ONIX is a small Python package that you should install from source.

The amazing work done with OpenMC has motivated and inspired our project with ONIX. I would like to take this opportunity to thank the many contributors of OpenMC and particularly Paul Romano who has been helping me a lot with OpenMC.

We hope you are going to find interest in ONIX and contribute to making it better. Please don’t hesitate to reach out to me if you have questions or if you want to contribute: j.detroullioud@gmail.com

Documentation: https://onix-documentation.readthedocs.io/en/latest/index.html
GitHub repository: https://github.com/jlanversin/ONIX
Article in Annals of Nuclear Energy: https://www.sciencedirect.com/science/article/pii/S0306454920306009?via%3Dihub


Julien de Troullioud de Lanversin


Thanks @Julien! Nice to see new efforts like this growing organically.


Hello, does it differ from openmc.deplete module ?

@azim Yes, ONIX is its own project that is maintained separately from OpenMC, whereas the openmc.deplete module is part of OpenMC proper. I would recommend reading @Julien’s paper on ONIX (linked above) to get a sense of what is possible with ONIX. There is also a paper on the openmc.deplete capabilities here for comparison.