fusion-energy / neutronics-workshopLinks
A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools
☆159Updated 3 weeks ago
Alternatives and similar repositories for neutronics-workshop
Users that are interested in neutronics-workshop are comparing it to the libraries listed below
Sorting:
- Native plotting GUI for model design and verification☆60Updated 3 weeks ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- Direct Accelerated Geometry Monte Carlo Toolkit☆116Updated 7 months ago
- List of open source projects related to OpenMC☆27Updated 3 months ago
- Python ENDF Parser☆36Updated last month
- Stochastic Calculator Of Neutron transport Equation☆51Updated 3 weeks ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆76Updated 2 months ago
- API for parsing and post-processing many MC simulations related files☆17Updated last month
- Tool for converting MCNP input files to OpenMC classes/XML☆45Updated last week
- ☆85Updated last year
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 2 months ago
- Tools to translate between different CSG geometry types☆39Updated 3 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 4 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 5 months ago
- High-Fidelity Multiphysics☆120Updated 2 weeks ago
- Sampling nuclear data and uncertainty☆52Updated this week
- Meshing library for nuclear workflows☆50Updated last week
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆39Updated 2 weeks ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆13Updated 4 months ago
- Convert OpenMC Geometry Components to CAD☆18Updated 9 months ago
- Example Jupyter notebooks for OpenMC☆38Updated 6 months ago
- Accelerated Discretized Geometry☆18Updated this week
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆43Updated last year
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated 3 weeks ago
- JADE, a novel nuclear data libraries V&V tool☆32Updated this week
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 3 weeks ago
- open-source Sn software☆42Updated this week
- Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.☆22Updated 4 months ago
- Nuclear data processing with legacy NJOY☆116Updated 3 weeks ago
- Collection of scripts for managing data for OpenMC☆34Updated 8 months ago