svalinn / DAGMC
Direct Accelerated Geometry Monte Carlo Toolkit
☆106Updated last week
Alternatives and similar repositories for DAGMC:
Users that are interested in DAGMC are comparing it to the libraries listed below
- Native plotting GUI for model design and verification☆55Updated last week
- ENRICO: Exascale Nuclear Reactor Investigative COde☆64Updated last year
- Nuclear data processing with legacy NJOY☆101Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆59Updated last month
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆53Updated 7 months ago
- NJOY for the 21st Century☆77Updated last year
- ☆73Updated 8 months ago
- Tools to translate between different CSG geometry types☆36Updated last year
- Tool for converting MCNP input files to OpenMC classes/XML☆37Updated last month
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated last week
- Collection of scripts for managing data for OpenMC☆29Updated 2 weeks ago
- Some Monte Carlo tools☆46Updated last month
- Sampling nuclear data and uncertainty☆49Updated 3 weeks ago
- Stochastic Calculator Of Neutron transport Equation☆39Updated this week
- Collection of models for reactor physics benchmarks☆50Updated 3 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆33Updated 3 weeks ago
- Example Jupyter notebooks for OpenMC☆35Updated 2 months ago
- List of open source projects related to OpenMC☆23Updated 5 months ago
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆34Updated this week
- A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutroni…☆132Updated this week
- Toolkit for reading and interacting with ENDF-6 formatted files☆36Updated 3 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆39Updated 7 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 5 months ago
- Monte Carlo Particle Lists☆30Updated 11 months ago
- A tool for making reproducible materials and standardizing use across several neutronics codes☆22Updated 3 months ago
- curated documents pertaining to early molten salt reactor research.☆21Updated 10 months ago
- NCrystal : a library for thermal neutron transport in crystals and other materials☆43Updated this week
- continuous energy monte carlo neutron transport in general geometries on GPUs☆43Updated 7 years ago