Opensource alternative for MCNP Monte Carlo simulation [closed]

Question

MCNP (Monte Carlo N-Particle) is a quite old but famous Fortran code to simulate radiation effects with Monte Carlo calculations. It can be used for radiation safety calculations.

It is very difficult and expensive to get the MCNP suite and the code is very hard to read.

What is an open source alternative for MCNP?

Answer 1

OpenMC (github)

The OpenMC project aims to provide a fully-featured Monte Carlo particle transport code based on modern methods. It is a constructive solid geometry, continuous-energy transport code that uses ACE format cross sections. The project started under the Computational Reactor Physics Group at MIT.

You can find more background information in this paper (ScienceDirect)