NCSU-NCSG / THORLinks
THOR is a radiation transport code for unstructured meshes.
☆16Updated last year
Alternatives and similar repositories for THOR
Users that are interested in THOR are comparing it to the libraries listed below
Sorting:
- Stochastic Calculator Of Neutron transport Equation☆53Updated 3 weeks ago
- Framework for REsearch in Nuclear ScIence and Engineering☆15Updated 4 years ago
- Collection of scripts for managing data for OpenMC☆38Updated 2 weeks ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆43Updated last year
- ENRICO: Exascale Nuclear Reactor Investigative COde☆66Updated 2 years ago
- High-Fidelity Multiphysics☆122Updated this week
- The Advanced Dimensional Depletion for Engineering of Reactors (ADDER) Software☆19Updated 3 weeks ago
- continuous energy monte carlo neutron transport in general geometries on GPUs☆46Updated 8 years ago
- An Open Nuclear Reactor Simulator and Reactor Core Analysis Tool☆28Updated 5 years ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆21Updated 6 months ago
- MC/DC: Monte Carlo Dynamic Code☆48Updated last week
- Collection of validation scripts, notebooks, results☆11Updated 2 weeks ago
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆63Updated 2 weeks ago
- Native plotting GUI for model design and verification☆64Updated last month
- Machine learning and artificial intelligence benchmarking library for nuclear engineering applications.☆29Updated 7 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆118Updated 10 months ago
- Repository for Moltres, a code for simulating Molten Salt Reactors☆76Updated last week
- A C++ Monte Carlo Neutron Transport Code☆20Updated last month
- A method of characteristics code for nuclear reactor physics calculations.☆175Updated last year
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆42Updated last year
- List of open source projects related to OpenMC☆30Updated 7 months ago
- Open-source Sn software☆53Updated 2 weeks ago
- Nuclear data processing with legacy NJOY☆123Updated 3 weeks ago
- Create parametric 3D fusion reactor CAD and neutronics models☆89Updated 4 months ago
- Sampling nuclear data and uncertainty☆54Updated 3 months ago
- Simulation of transmutation and decay in nuclear systems☆33Updated last week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 3 months ago
- Example Jupyter notebooks for OpenMC☆40Updated 10 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆49Updated this week
- Workflow and Template Toolkit for Simulation (WATTS)☆33Updated 6 months ago