ukaea / openmc_workshop
A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
☆39Updated 3 years ago
Alternatives and similar repositories for openmc_workshop
Users that are interested in openmc_workshop are comparing it to the libraries listed below
Sorting:
- ☆22Updated last week
- List of open source projects related to OpenMC☆24Updated 8 months ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 2 weeks ago
- Python ENDF Parser☆33Updated 9 months ago
- API for parsing and post-processing many MC simulations related files☆15Updated 3 weeks ago
- Tool for converting MCNP input files to OpenMC classes/XML☆41Updated 2 months ago
- Tools to translate between different CSG geometry types☆37Updated last year
- Accelerated Discretized Geometry☆14Updated this week
- Native plotting GUI for model design and verification☆57Updated 3 months ago
- Stochastic Calculator Of Neutron transport Equation☆43Updated this week
- JADE, a novel nuclear data libraries V&V tool☆26Updated this week
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆33Updated last month
- Collection of scripts for managing data for OpenMC☆31Updated 3 months 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
- Convert OpenMC Geometry Components to CAD☆17Updated 5 months ago
- A Python package for parsing FISPACT-II output☆28Updated 3 weeks ago
- ☆30Updated 2 months ago
- ENDF/B pre-processing codes (PREPRO)☆19Updated last year
- Sampling nuclear data and uncertainty☆51Updated last week
- ☆77Updated 11 months ago
- Simulate PWR,CANDU and VVER reactor (Assembly Model) using OpenMC https://openmc.readthedocs.io/en/stable/☆12Updated 5 years ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆69Updated this week
- Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis☆14Updated 4 years ago
- ENDF-6 data interface and nuclear data evaluation assist code☆16Updated 2 weeks ago
- OpenMC Monte Carlo Code☆17Updated 3 months ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆57Updated 3 weeks ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated last week
- Combines open source packages to produce an automated fusion specific neutronics workflow☆14Updated 2 years ago