sandialabs / MatMCNP
A utility code for generating material cards for MCNP
☆20Updated 3 years ago
Related projects ⓘ
Alternatives and complementary repositories for MatMCNP
- Python Interface and Utility Functions for MCNP6☆18Updated 7 years ago
- a companion for writing MCNP input decks☆11Updated 3 years ago
- Some Monte Carlo tools☆45Updated 2 weeks ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆57Updated 2 weeks ago
- ☆67Updated 5 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆32Updated this week
- Tool for converting MCNP input files to OpenMC classes/XML☆36Updated 2 weeks ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆34Updated last week
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago
- ENDF/B pre-processing codes (PREPRO)☆16Updated last year
- Sampling nuclear data and uncertainty☆48Updated last month
- List of open source projects related to OpenMC☆20Updated 2 months ago
- Tool to rename cells, surfaces, materials and universes in MCNP input files.☆11Updated last year
- Native plotting GUI for model design and verification☆47Updated 2 weeks ago
- Tools to translate between different CSG geometry types☆35Updated last year
- OpenMC Monte Carlo Code☆14Updated 2 months ago
- Collection of scripts for managing data for OpenMC☆27Updated 11 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆12Updated 5 months ago
- Monte Carlo Particle Lists☆29Updated 8 months ago
- Contains slides and example problems for OpenMC workshops and presentations☆23Updated 2 years ago
- A visualizer based on POVRAY to create visualizations of MCNPX input files without the need for MCNPX (created by Nick Michiels.)☆12Updated 12 years ago
- Python ENDF Parser☆26Updated 3 months ago
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆13Updated 7 years ago
- A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles☆10Updated 5 months ago
- Collection of models for reactor physics benchmarks☆48Updated last month
- Convert MCNP input files to a general CAD format☆34Updated 7 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆24Updated 2 months ago
- A tool for making reproducible materials and standardizing use across several neutronics codes☆20Updated last month
- NCrystal : a library for thermal neutron transport in crystals and other materials☆40Updated this week