xdg-org / xdgLinks
Accelerated Discretized Geometry
☆18Updated last week
Alternatives and similar repositories for xdg
Users that are interested in xdg are comparing it to the libraries listed below
Sorting:
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆17Updated 3 weeks ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 4 months ago
- Convert OpenMC Geometry Components to CAD☆20Updated 11 months ago
- List of open source projects related to OpenMC☆29Updated 5 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 3 years ago
- API for parsing and post-processing many MC simulations related files☆17Updated 2 weeks ago
- Computational inputs, reference outputs and experimental data for V&V systems☆15Updated 2 months ago
- Repository for running SINBAD benchmarks with OpenMC and UQ☆10Updated 3 years ago
- Python ENDF Parser☆39Updated 3 weeks ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- Tool for converting MCNP input files to OpenMC classes/XML☆50Updated 3 weeks ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆14Updated 6 months ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆80Updated 2 weeks ago
- open-source Sn software☆49Updated this week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- Native plotting GUI for model design and verification☆64Updated 3 weeks ago
- Tools to translate between different CSG geometry types☆40Updated 5 months ago
- Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.☆24Updated 6 months ago
- A Python package for parsing FISPACT-II output. https://deepwiki.com/fispact/pypact☆32Updated 3 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆40Updated last month
- Stochastic Calculator Of Neutron transport Equation☆51Updated 3 weeks ago
- ☆32Updated 3 months ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated last month
- OpenMC Monte Carlo Code☆23Updated 9 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆43Updated last year
- JADE, a novel nuclear data libraries V&V tool☆33Updated 3 weeks ago
- Sampling nuclear data and uncertainty☆53Updated last month
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆34Updated 4 months ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆31Updated 3 weeks ago
- ENDF-6 data interface and nuclear data evaluation assist code☆16Updated 6 months ago