Nuclear data processing with ACEMAKER, FRENDY and NJOY
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Updated
Apr 9, 2023 - Jupyter Notebook
Nuclear data processing with ACEMAKER, FRENDY and NJOY
OpenMC model of the EVOL reference MSFR. Includes code used to simulate and plot results in DTU student project "Impact of temperature feedback on reactivity parameters in the Molten Salt Fast Reactor" by Morten Nygaard (ongoing per June 2024). See "README.md" for detailed description of code and project.
Radiation transport simulations with a human phantoms in OpenMC. Meshes are created in Cubit Coreform. The phantom used in this project is obtained from: https://www.icrp.org/publication.asp?id=ICRP%20Publication%20145
Student research repository for independent research of the material attractiveness of prospective fuel to be used in microreactors
This repository contains a beginner-friendly tutorial for OpenMC, a Monte Carlo particle transport simulation code widely used in nuclear engineering.
Code to simulate the flux through cylinders with the MUTR Neutron Imager
The jupyter notebook contains python code which creates a BWR square assembly with a 3-by-3 fuel-less center.
openmc msr depletion capabilities example
A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC
A Python package that finds and converts OpenMC tally units.
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
Package Manager for Nuclear Engineering Development
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