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Title: 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:
; ; ; ; ; ; ;  [1]
  1. 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}
}
  • Purpose: The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. Methods: The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values.more » Results: For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. Conclusions: For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.« less
  • 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 MCNP4B, 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 beenmore » 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.« less
  • The aim of this work was to study the dosimetric potential of the Monte Carlo code MCNPX applied to the protontherapy field. For series of clinical configurations a comparison between simulated and experimental data was carried out, using the proton beam line of the MEDICYC isochronous cyclotron installed in the Centre Antoine Lacassagne in Nice. The dosimetric quantities tested were depth-dose distributions, output factors, and monitor units. For each parameter, the simulation reproduced accurately the experiment, which attests the quality of the choices made both in the geometrical description and in the physics parameters for beam definition. These encouraging resultsmore » enable us today to consider a simplification of quality control measurements in the future. Monitor Units calculation is planned to be carried out with preestablished Monte Carlo simulation data. The measurement, which was until now our main patient dose calibration system, will be progressively replaced by computation based on the MCNPX code. This determination of Monitor Units will be controlled by an independent semi-empirical calculation.« less
  • The long term human exploration goals that NASA has embraced, requires the need to understand the primary radiation and secondary particle production under a variety of environmental conditions. In order to perform accurate transport simulations for the incident particles found in the space environment, accurate nucleus-nucleus inelastic event generators are needed, and NASA is funding their development. For the first time, NASA is including the radiation problem into the . design of the next manned exploration vehicle. The NASA-funded FLUER-S (FLUKA Executing Under ROOT-Space) project has several goals beyond the improvement of the internal nuclear physics simulations. These include makingmore » FLUKA more user-friendly. Several tools have been developed to simplify the use of FLUKA without compromising its accuracy or versatility. Among these tools are a general source input, ability of distributive computing, simplification of geometry input, geometry and event visualization, and standard FLUKA scoring output analysis using a ROOT GUI. In addition to describing these tools we will show how they have been used for space radiation environment data analysis in MARIE, IVCPDS, and EVCPDS. Similar analyses can be performed for future radiation measurement detectors before they are deployed in order to optimize their design. These tools can also be used in the design of nuclear-based power systems on manned exploration vehicles and planetary surfaces. In addition to these space applications, the simulations are being used to support accelerator based experiments like the cross-section measurements being performed at HIMAC and NSRL at BNL.« less
  • Calculations have been performed using the Monte Carlo code, MORSE-CG, to determine the neutron streaming through various straight and stepped gaps between radiation shield sectors in the conceptual tokamak fusion power plant design STARFIRE. This design calls for ''pie-shaped'' radiation shields with gaps between segments. It is apparent that some type of offset, or stepped gap, configuration will be necessary to reduce neutron streaming through these gaps. To evaluate this streaming problem, a MORSE-to-MORSE coupling technique was used, consisting of two separate transport calculations, which together defined the entire transport problem. The results define the effectiveness of various gap configurationsmore » to eliminate radiation streaming.« less