njoy / ENDFtk
Toolkit for reading and interacting with ENDF-6 formatted files
☆36Updated 3 months ago
Alternatives and similar repositories for ENDFtk:
Users that are interested in ENDFtk are comparing it to the libraries listed below
- Tool for converting MCNP input files to OpenMC classes/XML☆37Updated last month
- ENDF/B pre-processing codes (PREPRO)☆16Updated last year
- Sampling nuclear data and uncertainty☆49Updated 3 weeks ago
- Python ENDF Parser☆27Updated 5 months ago
- List of open source projects related to OpenMC☆23Updated 5 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- ENDF-6 data interface and nuclear data evaluation assist code☆14Updated 3 weeks ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆59Updated last month
- ☆18Updated 5 months ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆53Updated 7 months ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆13Updated 8 months ago
- JADE, a novel nuclear data libraries V&V tool☆25Updated this week
- Monte Carlo Particle Lists☆30Updated 11 months ago
- Toolkit for working with ACE-formatted data files☆27Updated 3 months ago
- OpenMC Monte Carlo Code☆17Updated 5 months ago
- Tools to translate between different CSG geometry types☆36Updated last year
- ☆73Updated 8 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆34Updated this week
- Convert OpenMC Geometry Components to CAD☆16Updated 2 months ago
- A Python package for parsing FISPACT-II output☆26Updated 11 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆33Updated 3 weeks ago
- Native plotting GUI for model design and verification☆55Updated last week
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Stochastic Calculator Of Neutron transport Equation☆39Updated this week
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆26Updated 3 weeks ago
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago
- Collection of scripts for managing data for OpenMC☆29Updated 2 weeks ago
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 5 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 8 months ago