fusion-energy / openmc_source_plotterLinks
A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles
☆12Updated last year
Alternatives and similar repositories for openmc_source_plotter
Users that are interested in openmc_source_plotter are comparing it to the libraries listed below
Sorting:
- ENDF/B pre-processing codes (PREPRO)☆21Updated 2 years ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated 3 weeks ago
- ☆16Updated last year
- Tool for converting MCNP input files to OpenMC classes/XML☆52Updated last week
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆14Updated 7 months ago
- Simulate PWR,CANDU and VVER reactor (Assembly Model) using OpenMC https://openmc.readthedocs.io/en/stable/☆13Updated 6 years ago
- List of open source projects related to OpenMC☆30Updated 6 months ago
- Native plotting GUI for model design and verification☆64Updated 2 weeks ago
- Sampling nuclear data and uncertainty☆53Updated 2 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- Collection of scripts for managing data for OpenMC☆38Updated last week
- Python ENDF Parser☆39Updated last month
- Example Jupyter notebooks for OpenMC☆40Updated 9 months ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆39Updated 5 months ago
- ☆87Updated last year
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆42Updated last year
- Some Monte Carlo tools☆52Updated 3 weeks ago
- Open-source Sn software☆52Updated last week
- Combines open source packages to produce an automated fusion specific neutronics workflow☆15Updated 3 years ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated last week
- Repository for running SINBAD benchmarks with OpenMC and UQ☆10Updated 3 years ago
- A tool for making reproducible materials and standardizing use across several neutronics codes☆25Updated last year
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆34Updated 5 months ago
- Toolkit for working with ACE-formatted data files☆30Updated 3 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 3 years ago
- Converts CAD file(s) such as STP and SAT to a h5m file compatible with DAGMC based simulations using the Cubit Python API☆11Updated 2 years ago
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 4 years ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆66Updated 2 years ago
- Parametric 3-D modeling toolset for stellarator fusion devices☆38Updated 2 weeks ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆118Updated 9 months ago