OpenMC Monte Carlo Code

Geant4 toolkit for the simulation of the passage of particles through matter - NIM A 506 (2003) 250-303

MC/DC: Monte Carlo Dynamic Code

Method of Characteristics neutral particle transport code for reactor physics written in Julia.

High performance cyclic ray tracing algorithm for neutron transport in Julia.

THOR is a radiation transport code for unstructured meshes.

An unstructured mesh library for automated method of characteristic mesh generation

Monte Carlo neutron transport in C99. GPU via BarraCUDA (AMD MI300X, NVIDIA RTX). ENDF/B-VII.1 nuclear data. Validated against ICSBEP benchmarks. Zero dependencies.

An open-source MOOSE-based radiation transport and fluid activation solver.

Energy-dependent neutron transport Monte Carlo implemented in Rust.

Unity DOTS/ECS nuclear reactor simulator with real-time neutron physics, fission modeling, and RBMK positive void coefficient dynamics.

Neutron transport solver using collision probability method

Solves the RTE using the MOC on tetraedral meshes

C library for loading, querying, visualizing, and converting MCNP and OpenMC CSG geometry models

One Dimensional Neutron Transport Solver

Isotropic Monte Carlo simulations examining thermal neutron statistics in water, lead, and graphite, including animated neutron history, delta tracking and spherical geometry adaptations.

GEANT4 simulation code for LCS gamma-ray sources and flat efficiency moderated He-3 counters array dedicated to photoneutron reaction studies

Python library for debugging and visualizing Constructive Solid Geometry (CSG) models for MCNP and OpenMC

SRT_repo is built when first learning how to perform the neutron transport calculating during an SRT in Tsinghua University. This repo is primitive but easy-taking for the new comer when you want to buld your COM program hand-by-hand.

One velocity monte carlo for the rod problem with advection-diffusion of DNPs