openmc-dev / openmc-ecosystem
List of open source projects related to OpenMC
☆24Updated 8 months ago
Alternatives and similar repositories for openmc-ecosystem
Users that are interested in openmc-ecosystem are comparing it to the libraries listed below
Sorting:
- ☆22Updated this week
- Python ENDF Parser☆33Updated 9 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 3 weeks ago
- Tool for converting MCNP input files to OpenMC classes/XML☆41Updated 3 months ago
- API for parsing and post-processing many MC simulations related files☆15Updated 3 weeks ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated last week
- Accelerated Discretized Geometry☆14Updated this week
- Tools to translate between different CSG geometry types☆37Updated last year
- Convert OpenMC Geometry Components to CAD☆17Updated 5 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆33Updated last month
- Computational inputs, reference outputs and experimental data for V&V systems☆12Updated last month
- Workflow and Template Toolkit for Simulation (WATTS)☆30Updated last week
- Native plotting GUI for model design and verification☆57Updated 3 months ago
- Stochastic Calculator Of Neutron transport Equation☆43Updated this week
- Sampling nuclear data and uncertainty☆51Updated last week
- ENDF/B pre-processing codes (PREPRO)☆19Updated last year
- JADE, a novel nuclear data libraries V&V tool☆27Updated this week
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆29Updated 3 weeks ago
- Collection of scripts for managing data for OpenMC☆31Updated 3 months ago
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated this week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆43Updated this week
- OpenMC Monte Carlo Code☆17Updated 3 months ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆69Updated last week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated last week
- open-source Sn software☆33Updated this week
- A Python package for parsing FISPACT-II output☆28Updated 3 weeks ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 10 months ago
- Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis☆14Updated 4 years ago