afedynitch / x4i3
This is a fork of the x4i Python interface to the EXFOR database, originally developed at Lawrence Livermore National Security.
☆11Updated 10 months ago
Alternatives and similar repositories for x4i3:
Users that are interested in x4i3 are comparing it to the libraries listed below
- Python ENDF Parser☆31Updated 7 months ago
- Python package for reading, writing, verifying and translating ENDF-6 formatted files☆15Updated this week
- ENDF-6 data interface and nuclear data evaluation assist code☆15Updated last week
- Toolkit for reading and interacting with ENDF-6 formatted files☆38Updated 2 weeks ago
- ☆19Updated last week
- ENDF/B pre-processing codes (PREPRO)☆18Updated last year
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated last month
- JADE, a novel nuclear data libraries V&V tool☆25Updated this week
- The nuclear reaction model code☆35Updated this week
- Sampling nuclear data and uncertainty☆51Updated this week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆56Updated last month
- List of open source projects related to OpenMC☆24Updated 7 months ago
- The Reduced-Order Scattering Emulator (rose) is a user-friendly software for building efficient surrogate models for nuclear scattering☆11Updated this week
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆29Updated 2 months ago
- An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.☆11Updated 6 months ago
- A Python package for parsing FISPACT-II output☆28Updated last year
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆63Updated this week
- Tool for converting MCNP input files to OpenMC classes/XML☆39Updated last month
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- NCrystal : a library for thermal neutron transport in crystals and other materials☆44Updated last week
- Easily access hydrogen transport properties☆15Updated 3 months ago
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- Format description for ACE nuclear data files☆21Updated 2 years ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 10 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆40Updated this week
- Convert OpenMC Geometry Components to CAD☆17Updated 4 months ago
- Stochastic Calculator Of Neutron transport Equation☆42Updated 2 months ago
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- ☆14Updated 3 years ago
- Toolkit for working with ACE-formatted data files☆28Updated 2 weeks ago