svalinn / ALARA
Simulation of transmutation and decay in nuclear systems
☆28Updated last month
Alternatives and similar repositories for ALARA:
Users that are interested in ALARA are comparing it to the libraries listed below
- A tool for making reproducible materials and standardizing use across several neutronics codes☆23Updated 4 months ago
- Monte Carlo Particle Lists☆30Updated 11 months ago
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated this week
- Collection of scripts for managing data for OpenMC☆29Updated last month
- Toolkit for reading and interacting with ENDF-6 formatted files☆36Updated last week
- OpenMC Monte Carlo Code☆17Updated 3 weeks ago
- Sampling nuclear data and uncertainty☆50Updated last month
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 5 months ago
- CGMF nuclear fission fragment de-excitation statistical code☆25Updated 10 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 7 months ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆32Updated 2 months ago
- Tool for converting MCNP input files to OpenMC classes/XML☆37Updated last week
- ENDF/B pre-processing codes (PREPRO)☆17Updated last year
- Native plotting GUI for model design and verification☆55Updated last month
- Collection of models for reactor physics benchmarks☆51Updated 4 months ago
- List of open source projects related to OpenMC☆24Updated 5 months ago
- Parametric 3-D CAD modeling toolset for stellarator fusion devices☆28Updated this week
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆14Updated 2 weeks ago
- MC/DC: Monte Carlo Dynamic Code☆29Updated this week
- Direct Accelerated Geometry Monte Carlo Toolkit☆109Updated last week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆60Updated this week
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 9 months ago
- Format description for ACE nuclear data files☆20Updated 2 years ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆64Updated last year
- Toolkit for working with ACE-formatted data files☆27Updated last week
- Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.☆20Updated 4 months ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆28Updated last month
- Stochastic Calculator Of Neutron transport Equation☆39Updated 3 weeks ago
- Meshing library for nuclear workflows☆46Updated 2 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆35Updated last week