fusion-energy / openmc-plasma-source
Creates a plasma source as an openmc.source object from input parameters that describe the plasma
☆33Updated 2 months ago
Alternatives and similar repositories for openmc-plasma-source:
Users that are interested in openmc-plasma-source are comparing it to the libraries listed below
- A tool for making reproducible materials and standardizing use across several neutronics codes☆23Updated 5 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 9 months ago
- ☆18Updated last week
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 6 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Parametric 3-D CAD modeling toolset for stellarator fusion devices☆28Updated this week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated this week
- List of open source projects related to OpenMC☆24Updated 6 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 8 months ago
- Python ENDF Parser☆28Updated 7 months ago
- Native plotting GUI for model design and verification☆55Updated last month
- JADE, a novel nuclear data libraries V&V tool☆25Updated this week
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- Convert OpenMC Geometry Components to CAD☆16Updated 3 months ago
- Collection of scripts for managing data for OpenMC☆29Updated last month
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆14Updated last month
- An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.☆11Updated 5 months ago
- Combines open source packages to produce an automated fusion specific neutronics workflow☆14Updated 2 years ago
- Create parametric 3D fusion reactor CAD and neutronics models☆74Updated 2 weeks ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆60Updated last week
- ☆13Updated 11 months ago
- Coupled hydrogen/tritium transport and heat transfer modelling using FEniCS☆98Updated this week
- A Python package for downloading h5 cross section files for use in OpenMC.☆12Updated last month
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆28Updated last month
- A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles☆10Updated 9 months ago
- A Python package for parsing FISPACT-II output☆26Updated last year
- Sampling nuclear data and uncertainty☆50Updated last month
- Easily access hydrogen transport properties☆15Updated 2 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago