OpenMC Monte Carlo Code

Main yt repository

A suite of parsers designed to make interacting with SERPENT output files simple and flawless

Tool for converting MCNP input files to OpenMC classes/XML

MC/DC: Monte Carlo Dynamic Code

Online reprocessing for molten salt reactors

Create a surrogate ANN/GBR/GPR/SVR model for regression of nuclear reactor power distribution

A Genetic Algorithm (GA) / Discrete Particle Swarm Optimization/ Hybrid (GA-PSO) for nuclear fuel optimization using ML surrogates (DNN, KNN, Random Forest, Ridge) and OpenMC. Optimizes fuel loading patterns for a target k-eff and minimal Power Peaking Factor (PPF).

The package for reading mcnp input in a pythonic way

Tools to work with MCNP models and results

DIF3D plugin to the ARMI nuclear reactor analysis framework

Open source Bateman Solver based on LSODE

A pretty viewer for XSM files generated by DRAGON/DONJON or APOLLO neutronic codes

Generalized machine learning data-set builder for nuclear cross-section models

Some code for my Physics of Fission Reactors model analysis reports, written in MATLAB and Python

This GUI and terminal based program attempts to analyze both qualitative and quantitative characteristics of nuclear reactors. Currently can be done in GUI or in terminal. GUI tkinter is unable to conduct analysis on qualitative characteristics and does not have a scroll bar (two things I was unable to complete).

Tally table is a simple GUI program which extracts user defined tallies from a MCNP output.

Python scripts for Spent Nuclear Fuel Analysis, using SCALE/TRITON output files.

An attempt at developing an optimized control rod design for a nuclear reactor using a genetic algorithm and perturbation theory of the simple neutron diffusion equation model.

Development of background radiation monitoring with IoT based device