Stars
Python package for processing nuclear data distribution in ACE and ENDF format into HDF5 format for use in OpenMC. Adapted from the openmc data repo.
Python package for reading, writing, verifying and translating ENDF-6 formatted files
A python-based graphical user interface for the ion-solid interaction code SDTrimSP
My attempt at creating an OpenMC depletion chain file with isomer branching ratios for more than (n,gamma) reactions. Data extracted from JENDL-5 library at 14MeV
This repository holds various models and benchmarks
MCNP6 source subroutine for simulating D-T neutrons in Ti-T Target
A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes
Nuclear Decay Data for Dosimetric Calculations from ICRP 107
Creating basic step 2D tally plots from MCNP outputs
Simulation Environment for Radiotherapy Applications (SERA) is a package to help create treatment plans for treating tumors with boron neutron capture therapy. SERA allows calculating the dose that…
Anthropic's Interactive Prompt Engineering Tutorial
Simulation of transmutation and decay in nuclear systems
PROCESS is a systems code at UKAEA that calculates in a self-consistent manner the parameters of a fusion power plant with a specified performance, ensuring that its operating limits are not violat…
A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles
Computational inputs, reference outputs and experimental data for V&V systems
Monte Carlo neutron source terms for SHINE's neutron generators
A collection of notebooks/recipes showcasing some fun and effective ways of using Claude.
ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers coupling with the open-source transport code OpenMC.
Tool to convert MCNP geometries into TRIPOLI-4 geometries.
Repository for FES Shutdown Dose Rate Workflows project
The Bateman equation in C++ and a Python wrapper thereof.