Alternatives and similar repositories for fudge

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

• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆26Updated last week
• eepeterson / openmc_fusion_benchmarks
  Automated V&V for fusion relevant CAD-based benchmarks
  ☆23Updated this week
• paulromano / endf-python
  Python ENDF Parser
  ☆33Updated 10 months ago
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆16Updated 3 weeks ago
• Fusion4Energy / F4Enix
  API for parsing and post-processing many MC simulations related files
  ☆17Updated this week
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆51Updated last month
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆42Updated 2 weeks ago
• pshriwise / xdg
  Accelerated Discretized Geometry
  ☆17Updated this week
• toshihikokawano / DeCE
  ENDF-6 data interface and nuclear data evaluation assist code
  ☆16Updated last month
• IAEA-NDS / PREPRO
  ENDF/B pre-processing codes (PREPRO)
  ☆19Updated last year
• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆39Updated 3 years ago
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆29Updated this week
• 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
• lanl / TOGA
  Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections
  ☆11Updated 2 years ago
• exasmr / openmc
  OpenMC Monte Carlo Code
  ☆18Updated 4 months ago
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆41Updated last week
• IAEA-NDS / open-benchmarks
  Computational inputs, reference outputs and experimental data for V&V systems
  ☆13Updated last month
• idaholab / MontePy
  MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
  ☆46Updated last week
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆18Updated 6 months ago
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆43Updated 3 weeks ago
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆37Updated last week
• fusion-energy / openmc_regular_mesh_plotter
  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
  ☆11Updated last month
• Open-Sn / opensn
  open-source Sn software
  ☆34Updated this week
• fispact / pypact
  A Python package for parsing FISPACT-II output
  ☆29Updated last week
• openmc-data-storage / isotope-xs-plotter
  XSPlot - Neutron cross section plotter for isotopes
  ☆14Updated last year
• lanl / mcnptools
  ☆79Updated last year
• openmc-dev / data
  Collection of scripts for managing data for OpenMC
  ☆33Updated 4 months ago
• KDSource / KDSource
  Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.
  ☆22Updated 3 weeks ago
• mctools / ncrystal
  NCrystal : a library for thermal neutron transport in crystals and other materials
  ☆46Updated this week
• CORE-GATECH-GROUP / serpent-tools
  A suite of parsers designed to make interacting with SERPENT output files simple and flawless
  ☆57Updated last month