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Title: Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.

Abstract

Version 01 US DOE 10CFR810 Jurisdiction. MCNP6™ is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5™ [X-503] and MCNPX™ [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering andmore » Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code PARTISN.« less

Authors:
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1272161
Report Number(s):
MCNP6.1/MCNP5/MCNPX-EXE; 004380MLTPL00
RSICC ID: C00810MNYCP
DOE Contract Number:
AC05-00OR22725
Resource Type:
Software
Software Revision:
00
Software Package Number:
004380
Software CPU:
MLTPL
Source Code Available:
Yes
Other Software Info:
Owner Installation: LOS ALAMOS NATIONAL LABORATORY Contributors: Los Alamos National Laboratory, Los Alamos, New Mexico. Export control regulations restrict the distribution of Fortran source code. If restrictions apply, RSICC will send the executable-only version. Please note that included MCNP6.1 executables run only on the machines listed. In addition to the code capabilities of MCNP5 and MCNPX, MCNP6 includes several significant new capabilities not found in either of the parent codes. See abstract for descriptions. KEYWORDS: COMPLEX GEOMETRY; COUPLED; CROSS SECTIONS; ELECTRON; GAMMA-RAY; MONTE CARLO; NEUTRON
Country of Publication:
United States

Citation Formats

GOORLEY, TIM. Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.. Computer software. Vers. 00. USDOE. 16 Jul. 2013. Web.
GOORLEY, TIM. (2013, July 16). Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries. (Version 00) [Computer software].
GOORLEY, TIM. Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.. Computer software. Version 00. July 16, 2013.
@misc{osti_1272161,
title = {Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries., Version 00},
author = {GOORLEY, TIM},
abstractNote = {Version 01 US DOE 10CFR810 Jurisdiction. MCNP6™ is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5™ [X-503] and MCNPX™ [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code PARTISN.},
doi = {},
year = {Tue Jul 16 00:00:00 EDT 2013},
month = {Tue Jul 16 00:00:00 EDT 2013},
note =
}

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  • Version 00 US DOE 10CFR810 Jurisdiction. MCNP6™ is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5™ [X-503] and MCNPX™ [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or newmore » tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code PARTISN.« less
  • Version 01 MCNP6™ is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP6.1.1Beta is a follow-on to the MCNP6.1 production version which itself was the culmination of a multi-year effort to merge the MCNP5™ [X-503] and MCNPX™ [PEL11] codes into a single product. This MCNP6.1.1 beta has been released in order to provide the radiation transport community with the latest feature developments and bug fixes in the code. MCNP6.1.1 has taken input from a group of people, residing in the Los Alamos National Laboratory's (LANL) X Computationalmore » Physics Division, Radiation Transport Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5). They have combined their code development efforts to produce this next evolution of MCNP. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams.« less
  • The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a selectmore » group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to calculate radiation dose due to the neutron environment around a MEA is shown. An uncertainty of a factor of three in the MEA calculations is shown to be due to uncertainties in the geometry modeling. It is believed that the methodology is sound and that good agreement between simulation and experiment has been demonstrated.« less
  • Version: 00 US DOE 10CFR810 Jurisdiction. The Monte Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photons from fission requires that particle emissions and interactions following a fission event be described as close to reality as possible. The -PoliMi extension to MCNP and to MCNPX was developed to simulate correlated-particle and the subsequent interactions as close as possible to the physical behavior. Initially, MCNP-PoliMi, a modification of MCNP4C, was developed. The first version was developed in 2001-2002 and released in early 2004 to the Radiation Safety Information Computational Center (RSICC). It was developed formore » research purposes, to simulate correlated counts in organic scintillation detectors, sensitive to fast neutrons and gamma rays. Originally, the field of application was nuclear safeguards; however subsequent improvements have enhanced the ability to model measurements in other research fields as well. During 2010-2011 the -PoliMi modification was ported into MCNPX-2.7.0, leading to the development of MCNPX-PoliMi. Now the -PoliMi v2.0 modifications are distributed as a patch to MCNPX-2.7.0 which currently is distributed in the RSICC PACKAGE BCC-004 MCNP6_BETA2/MCNP5/MCNPX. Also included in the package is MPPost, a versatile code that provides simulated detector response. By taking advantage of the modifications in MCNPX-PoliMi, MPPost can provide an accurate simulation of the detector response for a variety of detection scenarios.« less
  • Version: 00 US DOE 10CFR810 Jurisdiction. The Monte Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photons from fission requires that particle emissions and interactions following a fission event be described as close to reality as possible. The -PoliMi extension to MCNP and to MCNPX was developed to simulate correlated-particle and the subsequent interactions as close as possible to the physical behavior. Initially, MCNP-PoliMi, a modification of MCNP4C, was developed. The first version was developed in 2001-2002 and released in early 2004 to the Radiation Safety Information Computational Center (RSICC). It was developed formore » research purposes, to simulate correlated counts in organic scintillation detectors, sensitive to fast neutrons and gamma rays. Originally, the field of application was nuclear safeguards; however subsequent improvements have enhanced the ability to model measurements in other research fields as well. During 2010-2011 the -PoliMi modification was ported into MCNPX-2.7.0, leading to the development of MCNPX-PoliMi. Now the -PoliMi v2.0 modifications are distributed as a patch to MCNPX-2.7.0 which currently is distributed in the RSICC PACKAGE BCC-004 MCNP6_BETA2/MCNP5/MCNPX. Also included in the package is MPPost, a versatile code that provides simulated detector response. By taking advantage of the modifications in MCNPX-PoliMi, MPPost can provide an accurate simulation of the detector response for a variety of detection scenarios.« less

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