NCSU-NCSG / THOR
THOR is a radiation transport code for unstructured meshes.
☆11Updated this week
Related projects: ⓘ
- Stochastic Calculator Of Neutron transport Equation☆34Updated last week
- Framework for REsearch in Nuclear ScIence and Engineering☆13Updated 3 years ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆17Updated last year
- ENRICO: Exascale Nuclear Reactor Investigative COde☆61Updated last year
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆35Updated 2 months ago
- Collection of scripts for managing data for OpenMC☆27Updated 9 months ago
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆50Updated this week
- The Advanced Dimensional Depletion for Engineering of Reactors (ADDER) Software☆13Updated 2 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 3 weeks ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆52Updated 3 months ago
- Native plotting GUI for model design and verification☆45Updated last month
- open-source Sn software☆18Updated this week
- Tools to translate between different CSG geometry types☆32Updated last year
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆24Updated 3 years ago
- Collection of models for reactor physics benchmarks☆48Updated 2 months ago
- Sampling nuclear data and uncertainty☆47Updated 2 weeks ago
- Monte Carlo Particle Lists☆28Updated 6 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- This repository holds codes from the book Computational Nuclear Engineering and Radiological Science with Python☆14Updated 4 years ago
- An Open Nuclear Reactor Simulator and Reactor Core Analysis Tool☆23Updated 3 years ago
- OpenMC wrapped as a Moose App☆17Updated last year
- List of open source projects related to OpenMC☆19Updated 3 weeks ago
- Format description for ACE nuclear data files☆19Updated 2 years ago
- OpenMC Monte Carlo Code☆14Updated 3 weeks ago
- ☆16Updated last month
- High-Fidelity Multiphysics☆89Updated last week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆30Updated this week
- Tool for converting MCNP input files to OpenMC classes/XML☆33Updated last month
- Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.☆20Updated 2 weeks ago
- A Python package for parsing FISPACT-II output☆25Updated 6 months ago