Alternatives and similar repositories for JADE:

Users that are interested in JADE are comparing it to the libraries listed below

• paulromano / endf-python
  Python ENDF Parser
  ☆27Updated 5 months ago
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆37Updated last month
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆13Updated 8 months ago
• eepeterson / openmc_fusion_benchmarks
  ☆18Updated 5 months ago
• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆39Updated 3 years ago
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆49Updated 3 weeks ago
• GEOUNED-org / GEOUNED
  A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes
  ☆59Updated last month
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆23Updated 5 months ago
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆36Updated 3 months ago
• fispact / pypact
  A Python package for parsing FISPACT-II output
  ☆26Updated 11 months ago
• watts-dev / watts
  Workflow and Template Toolkit for Simulation (WATTS)
  ☆25Updated 5 months ago
• mctools / mcpl
  Monte Carlo Particle Lists
  ☆30Updated 11 months ago
• CORE-GATECH-GROUP / serpent-tools
  A suite of parsers designed to make interacting with SERPENT output files simple and flawless
  ☆53Updated 7 months ago
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆36Updated last year
• IAEA-NDS / PREPRO
  ENDF/B pre-processing codes (PREPRO)
  ☆16Updated last year
• idaholab / MontePy
  MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
  ☆34Updated this week
• fusion-energy / openmc_regular_mesh_plotter
  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
  ☆11Updated 8 months ago
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆16Updated 2 months ago
• highness-eu / NJOY-NCrystal-Library
  Thermal Scattering Library created with NJOY+NCrystal
  ☆9Updated 2 years ago
• lanl / TOGA
  Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections
  ☆10Updated 2 years ago
• aurora-multiphysics / aurora
  Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.
  ☆39Updated 7 months ago
• ezsolti / RFP
  ☆29Updated last year
• toshihikokawano / DeCE
  ENDF-6 data interface and nuclear data evaluation assist code
  ☆14Updated 3 weeks ago
• openmc-dev / data
  Collection of scripts for managing data for OpenMC
  ☆29Updated 2 weeks ago
• ukaea / neutronics_material_maker
  A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak
  ☆19Updated 3 years ago
• openmc-dev / plotter
  Native plotting GUI for model design and verification
  ☆55Updated last week
• jlanversin / ONIX
  ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…
  ☆39Updated last month
• lanl / mcnptools
  ☆73Updated 8 months ago
• mit-crpg / BEAVRS
  A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant
  ☆33Updated 3 weeks ago
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆39Updated this week