ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers coupling with the open-source transport code OpenMC.
☆42Dec 11, 2024Updated last year
Alternatives and similar repositories for ONIX
Users that are interested in ONIX are comparing it to the libraries listed below
Sorting:
- Tool for converting MCNP input files to OpenMC classes/XML☆53Dec 9, 2025Updated 2 months ago
- Format description for ACE nuclear data files☆22Apr 28, 2022Updated 3 years ago
- List of open source projects related to OpenMC☆31Updated this week
- OpenMC wrapped as a Moose App☆16May 1, 2023Updated 2 years ago
- Workflow and Template Toolkit for Simulation (WATTS)☆33Jun 26, 2025Updated 8 months ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆66Jul 21, 2023Updated 2 years ago
- Simulation of transmutation and decay in nuclear systems☆35Feb 21, 2026Updated last week
- Collection of validation scripts, notebooks, results☆11Dec 23, 2025Updated 2 months ago
- MC/DC: Monte Carlo Dynamic Code☆52Updated this week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆62Sep 30, 2025Updated 5 months ago
- Native plotting GUI for model design and verification☆65Feb 9, 2026Updated 3 weeks ago
- Python ENDF Parser☆40Feb 12, 2026Updated 2 weeks ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆32Jan 29, 2026Updated last month
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆28May 19, 2021Updated 4 years ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Sep 13, 2021Updated 4 years ago
- ENDF/B pre-processing codes (PREPRO)☆22Aug 14, 2023Updated 2 years ago
- Papillon Nuclear Data Library☆19Jul 18, 2025Updated 7 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Oct 20, 2022Updated 3 years ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆42Jan 23, 2026Updated last month
- NCrystal : a library for thermal neutron transport in crystals and other materials☆47Updated this week
- Repository for running SINBAD benchmarks with OpenMC and UQ☆10May 18, 2022Updated 3 years ago
- A code package to produce ACE-formatted files for MCNP calculations.☆14Jun 14, 2025Updated 8 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆123Feb 16, 2026Updated 2 weeks ago
- Collection of scripts for managing data for OpenMC☆39Dec 23, 2025Updated 2 months ago
- OpenMC Monte Carlo Code☆24Jan 31, 2025Updated last year
- Repository for Moltres, a code for simulating Molten Salt Reactors☆78Feb 9, 2026Updated 3 weeks ago
- Tools to translate between different CSG geometry types☆41May 28, 2025Updated 9 months ago
- Simulate PWR,CANDU and VVER reactor (Assembly Model) using OpenMC https://openmc.readthedocs.io/en/stable/☆14Nov 29, 2019Updated 6 years ago
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆13Jul 30, 2017Updated 8 years ago
- Sampling nuclear data and uncertainty☆56Updated this week
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆43Jun 25, 2024Updated last year
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆54Updated this week
- CGMF nuclear fission fragment de-excitation statistical code☆28Mar 26, 2024Updated last year
- A tool for making reproducible materials and standardizing use across several neutronics codes☆25Oct 1, 2024Updated last year
- Example Jupyter notebooks for OpenMC☆44Feb 28, 2025Updated last year
- ☆33Jul 21, 2025Updated 7 months ago
- continuous energy monte carlo neutron transport in general geometries on GPUs☆46Jun 8, 2017Updated 8 years ago
- An open-source nuclear reactor analysis automation framework that helps design teams increase efficiency and quality☆262Updated this week
- High-Fidelity Multiphysics☆128Feb 19, 2026Updated last week