fusion-energy / openmc_source_plotterLinks
A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles
☆10Updated last year
Alternatives and similar repositories for openmc_source_plotter
Users that are interested in openmc_source_plotter are comparing it to the libraries listed below
Sorting:
- ☆16Updated last year
- ENDF/B pre-processing codes (PREPRO)☆19Updated last year
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated last week
- Combines open source packages to produce an automated fusion specific neutronics workflow☆15Updated 3 years ago
- Native plotting GUI for model design and verification☆60Updated 6 months ago
- Tool for converting MCNP input files to OpenMC classes/XML☆44Updated 2 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆12Updated 3 months ago
- List of open source projects related to OpenMC☆26Updated 2 months ago
- Sampling nuclear data and uncertainty☆52Updated last month
- Parametric 3-D CAD modeling toolset for stellarator fusion devices☆34Updated this week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆33Updated last month
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- open-source Sn software☆37Updated this week
- Simulate PWR,CANDU and VVER reactor (Assembly Model) using OpenMC https://openmc.readthedocs.io/en/stable/☆12Updated 5 years ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆36Updated last month
- Python ENDF Parser☆34Updated last year
- Simulation of transmutation and decay in nuclear systems☆29Updated 6 months ago
- Some Monte Carlo tools☆50Updated 3 weeks ago
- ☆83Updated last year
- Collection of scripts for managing data for OpenMC☆33Updated 6 months ago
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆41Updated 7 months ago
- Example Jupyter notebooks for OpenMC☆37Updated 5 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆42Updated last year
- Converts CAD file(s) such as STP and SAT to a h5m file compatible with DAGMC based simulations using the Cubit Python API☆12Updated last year
- A visualizer based on POVRAY to create visualizations of MCNPX input files without the need for MCNPX (created by Nick Michiels.)☆13Updated 12 years ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 3 weeks ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- Monte Carlo Particle Lists☆32Updated this week
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 5 months ago
- a companion for writing MCNP input decks☆12Updated 4 years ago