The MCNPX Monte Carlo Radiation Transport Code

Abstract

MCNPX (Monte Carlo N‐Particle eXtended) is a general‐purpose Monte Carlo radiation transport code with three‐dimensional geometry and continuous‐energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi‐processing computer platforms. MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low‐energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.