IAEA-NDS / PREPRO
ENDF/B pre-processing codes (PREPRO)
☆18Updated last year
Alternatives and similar repositories for PREPRO:
Users that are interested in PREPRO are comparing it to the libraries listed below
- Toolkit for reading and interacting with ENDF-6 formatted files☆39Updated this week
- Tool for converting MCNP input files to OpenMC classes/XML☆40Updated last month
- Python ENDF Parser☆31Updated 8 months ago
- List of open source projects related to OpenMC☆24Updated 7 months ago
- Sampling nuclear data and uncertainty☆51Updated last week
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- JADE, a novel nuclear data libraries V&V tool☆25Updated this week
- Toolkit for working with ACE-formatted data files☆28Updated 2 weeks ago
- ☆19Updated last week
- ENDF-6 data interface and nuclear data evaluation assist code☆16Updated this week
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- Collection of scripts for managing data for OpenMC☆30Updated 2 months ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆29Updated 2 months ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 2 months ago
- NCrystal : a library for thermal neutron transport in crystals and other materials☆44Updated this week
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Native plotting GUI for model design and verification☆57Updated 2 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆40Updated this week
- Stochastic Calculator Of Neutron transport Equation☆42Updated 2 months ago
- A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles☆10Updated 10 months ago
- Format description for ACE nuclear data files☆21Updated 2 years ago
- Convert OpenMC Geometry Components to CAD☆17Updated 4 months ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆63Updated this week
- Tools to translate between different CSG geometry types☆36Updated last year
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆56Updated last month
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆39Updated 3 months ago
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- Workflow and Template Toolkit for Simulation (WATTS)☆29Updated 7 months ago
- OpenMC Monte Carlo Code☆17Updated 2 months ago