jlball / arc-nonproliferation
An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.
☆10Updated 2 weeks ago
Related projects: ⓘ
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆12Updated 3 months ago
- ☆16Updated last month
- Easily access hydrogen transport properties☆13Updated last week
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A Python package for downloading h5 cross section files for use in OpenMC.☆10Updated last year
- Python ENDF Parser☆25Updated last month
- ☆27Updated last year
- List of open source projects related to OpenMC☆19Updated 3 weeks ago
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated last year
- A Python package for parsing FISPACT-II output☆25Updated 6 months ago
- Learn how to simulate hydrogen transport with FESTIM☆15Updated this week
- Meshing library for nuclear workflows☆41Updated last month
- OpenMC Monte Carlo Code☆14Updated 3 weeks ago
- Coupled hydrogen/tritium transport and heat transfer modelling using FEniCS☆88Updated last week
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 3 weeks ago
- Stochastic Calculator Of Neutron transport Equation☆34Updated last week
- Tool to estimate H retention in tokamak divertors☆10Updated 2 years ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆30Updated this week
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆26Updated last month
- open-source Sn software☆18Updated this week
- A tool for making reproducible materials and standardizing use across several neutronics codes☆18Updated 2 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆35Updated 2 months ago
- JADE, a novel nuclear data libraries V&V tool☆23Updated last month
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆53Updated last month
- Tool for converting MCNP input files to OpenMC classes/XML☆33Updated last month
- Native plotting GUI for model design and verification☆45Updated last month
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 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☆12Updated 11 months ago
- Combines open source packages to produce an automated fusion specific neutronics workflow☆13Updated 2 years ago