# 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.

- Authors:

- Los Alamos National Laboratory, MS K575, Los Alamos, New Mexico 87545 (United States)

- Publication Date:

- OSTI Identifier:
- 21055006

- Resource Type:
- Journal Article

- Resource Relation:
- Journal Name: AIP Conference Proceedings; Journal Volume: 896; Journal Issue: 1; Conference: Hadronic shower simulation workshop, Batavia, IL (United States), 6-8 Sep 2006; Other Information: DOI: 10.1063/1.2720459; (c) 2007 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA)

- Country of Publication:
- United States

- Language:
- English

- Subject:
- 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; ELECTRONS; ENERGY ABSORPTION; HEAVY ION REACTIONS; HEAVY IONS; INTERACTIVE DISPLAY DEVICES; LIGHT IONS; M CODES; MONTE CARLO METHOD; NEUTRON TRANSPORT THEORY; NEUTRONS; NUCLEAR DATA COLLECTIONS; PARTICLE DECAY; PARTICLE INTERACTIONS; PHOTONS; RADIATION TRANSPORT; THREE-DIMENSIONAL CALCULATIONS; TRANSMUTATION

### Citation Formats

```
Waters, Laurie S., McKinney, Gregg W., Durkee, Joe W., Fensin, Michael L., Hendricks, John S., James, Michael R., Johns, Russell C., and Pelowitz, Denise B..
```*The MCNPX Monte Carlo Radiation Transport Code*. United States: N. p., 2007.
Web. doi:10.1063/1.2720459.

```
Waters, Laurie S., McKinney, Gregg W., Durkee, Joe W., Fensin, Michael L., Hendricks, John S., James, Michael R., Johns, Russell C., & Pelowitz, Denise B..
```*The MCNPX Monte Carlo Radiation Transport Code*. United States. doi:10.1063/1.2720459.

```
Waters, Laurie S., McKinney, Gregg W., Durkee, Joe W., Fensin, Michael L., Hendricks, John S., James, Michael R., Johns, Russell C., and Pelowitz, Denise B.. Mon .
"The MCNPX Monte Carlo Radiation Transport Code". United States.
doi:10.1063/1.2720459.
```

```
@article{osti_21055006,
```

title = {The MCNPX Monte Carlo Radiation Transport Code},

author = {Waters, Laurie S. and McKinney, Gregg W. and Durkee, Joe W. and Fensin, Michael L. and Hendricks, John S. and James, Michael R. and Johns, Russell C. and Pelowitz, Denise B.},

abstractNote = {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.},

doi = {10.1063/1.2720459},

journal = {AIP Conference Proceedings},

number = 1,

volume = 896,

place = {United States},

year = {Mon Mar 19 00:00:00 EDT 2007},

month = {Mon Mar 19 00:00:00 EDT 2007}

}