Alternatives and similar repositories for isotope-xs-plotter

Users that are interested in isotope-xs-plotter are comparing it to the libraries listed below

• AnderGray / openmc_uq_sinbad
  Repository for running SINBAD benchmarks with OpenMC and UQ
  ☆10Updated 3 years ago
• lanl / TOGA
  Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections
  ☆11Updated 3 years ago
• eepeterson / openmc_fusion_benchmarks
  Automated V&V for fusion relevant CAD-based benchmarks
  ☆23Updated last week
• jlball / arc-nonproliferation
  An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.
  ☆11Updated last year
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆18Updated last month
• xdg-org / xdg
  Accelerated Discretized Geometry
  ☆22Updated last week
• openmc-data-storage / openmc_data_downloader
  A Python package for downloading h5 cross section files for use in OpenMC.
  ☆15Updated last year
• paulromano / endf-python
  Python ENDF Parser
  ☆39Updated 3 months ago
• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆41Updated 4 years ago
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆55Updated 3 months ago
• llnl / fudge
  For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data
  ☆32Updated 3 months ago
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆30Updated 7 months ago
• ukaea / neutronics_material_maker
  A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak
  ☆19Updated 4 years ago
• aurora-multiphysics / aurora
  Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.
  ☆43Updated last year
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆55Updated last week
• ezsolti / RFP
  ☆33Updated 6 months ago
• Fusion4Energy / F4Enix
  API for parsing and post-processing many MC simulations related files
  ☆17Updated this week
• fusion-energy / openmc_regular_mesh_plotter
  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
  ☆14Updated 8 months ago
• Open-Sn / opensn
  Open-source Sn software
  ☆54Updated this week
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆40Updated 7 months ago
• exasmr / openmc
  OpenMC Monte Carlo Code
  ☆24Updated 11 months ago
• idaholab / MontePy
  MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
  ☆50Updated this week
• openmsr / CAD_to_OpenMC
  code to target the conversion from a step-file to a h5m-geometry for neutronics
  ☆36Updated 6 months ago
• GEOUNED-org / GEOUNED
  A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes
  ☆85Updated last month
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆20Updated last year
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆52Updated last month
• fusion-energy / openmc-plasma-source
  Creates a plasma source as an openmc.source object from input parameters that describe the plasma
  ☆39Updated 7 months ago
• openmc-dev / plotter
  Native plotting GUI for model design and verification
  ☆64Updated this week
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆34Updated last week
• IAEA-NDS / open-benchmarks
  Computational inputs, reference outputs and experimental data for V&V systems
  ☆15Updated 4 months ago