ONIX: An open-source depletion code

Highlights: • New open-source depletion software for nuclear reactor physics simulations. • Provides coupling with open-source transport code OpenMC as well as standalone capability. • Provides developer-friendly Python interface, highly flexible nuclear data libraries processing, and output data postprocessing. • First depletion software adapted for nuclear archaeology and nuclear forensic analysis. • Software validated through comparisons with various numerical and experimental benchmarks. Open Source software enables innovative, community-based software development. ONIX brings this concept to the field of depletion calculations. It is an open-source depletion software to be used for nuclear reactor simulations, for fissile material production analysis as well as for nuclear arms control applications. ONIX provides a module to solve the depletion equation using a Chebyshev Rational Approximation Method. For the generation of one-group cross sections, it includes a coupling interface for the open-source neutron transport code, OpenMC, as well as a module to read pre-computed values in a stand-alone mode. ONIX has special features to optimize nuclear data libraries, to update isomeric branching ratio during burnup, and to support automation of simulations for nuclear archaeology. ONIX has been validated against results from numerical and experimental benchmarks, and its results agree with other methods within expected error ranges.