eepeterson / openmc_fusion_benchmarks
☆18Updated this week
Alternatives and similar repositories for openmc_fusion_benchmarks:
Users that are interested in openmc_fusion_benchmarks are comparing it to the libraries listed below
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated last month
- List of open source projects related to OpenMC☆24Updated 6 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 10 months ago
- Python ENDF Parser☆31Updated 7 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Accelerated Discretized Geometry☆13Updated this week
- Openmc-FEnicsx for muLtiphysics tutorIAl☆14Updated 7 months ago
- An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.☆11Updated 6 months ago
- Convert OpenMC Geometry Components to CAD☆17Updated 4 months ago
- Easily access hydrogen transport properties☆15Updated 3 months ago
- Tool for converting MCNP input files to OpenMC classes/XML☆39Updated last month
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- JADE, a novel nuclear data libraries V&V tool☆25Updated last week
- A Python package for parsing FISPACT-II output☆28Updated last year
- open-source Sn software☆27Updated this week
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- OpenMC Monte Carlo Code☆17Updated last month
- A Python package for downloading h5 cross section files for use in OpenMC.☆12Updated 2 months ago
- A tool for making reproducible materials and standardizing use across several neutronics codes☆23Updated 5 months ago
- Native plotting GUI for model design and verification☆56Updated 2 months ago
- ENDF-6 data interface and nuclear data evaluation assist code☆15Updated last week
- Stochastic Calculator Of Neutron transport Equation☆42Updated 2 months ago
- ☆30Updated last month
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆29Updated 2 months ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆33Updated 3 months ago
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Sampling nuclear data and uncertainty☆51Updated last week
- Toolkit for reading and interacting with ENDF-6 formatted files☆38Updated 2 weeks ago
- Tools to translate between different CSG geometry types☆36Updated last year
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 9 months ago