ONIX: An open-source depletion code

J. de Troullioud de Lanversin, M. Kütt, A. Glaser

Research output: Contribution to journalArticlepeer-review

Abstract

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.

Original languageEnglish (US)
Article number107903
JournalAnnals of Nuclear Energy
Volume151
DOIs
StatePublished - Feb 2021

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering

Keywords

  • Burnup
  • Depletion
  • ONIX
  • Open source
  • OpenMC
  • Validation

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