Monte Carlo NParticle Transport Code System Including MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries.
Abstract
Version 01 US DOE 10CFR810 Jurisdiction. MCNP6™ is a generalpurpose, continuousenergy, generalizedgeometry, timedependent, Monte Carlo radiationtransport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multiyear effort to merge the MCNP5™ [X503] 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 crosssection table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variancereduction 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 (XCP3), and Nuclear Engineering andmore »
 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/MCNPXEXE; 004380MLTPL00
RSICC ID: C00810MNYCP
 DOE Contract Number:
 AC0500OR22725
 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 executableonly 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; GAMMARAY; MONTE CARLO; NEUTRON
 Country of Publication:
 United States
Citation Formats
GOORLEY, TIM. Monte Carlo NParticle Transport Code System Including MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries..
Computer software. Vers. 00. USDOE. 16 Jul. 2013.
Web.
GOORLEY, TIM. (2013, July 16). Monte Carlo NParticle Transport Code System Including MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries. (Version 00) [Computer software].
GOORLEY, TIM. Monte Carlo NParticle Transport Code System Including MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries..
Computer software. Version 00. July 16, 2013.
@misc{osti_1272161,
title = {Monte Carlo NParticle Transport Code System Including MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries., Version 00},
author = {GOORLEY, TIM},
abstractNote = {Version 01 US DOE 10CFR810 Jurisdiction. MCNP6™ is a generalpurpose, continuousenergy, generalizedgeometry, timedependent, Monte Carlo radiationtransport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multiyear effort to merge the MCNP5™ [X503] 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 crosssection table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variancereduction 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 (XCP3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN5, formerly D5), 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 =
}

Version 00 US DOE 10CFR810 Jurisdiction. MCNP6™ is a generalpurpose, continuousenergy, generalizedgeometry, timedependent, Monte Carlo radiationtransport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multiyear effort to merge the MCNP5™ [X503] 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 crosssection table range, a material burnup feature, and delayed particle production. Expanded and/or newmore »

Monte Carlo NParticle Transport Code System Including MCNP6.1.1BETA, MCNP6.1, MCNP51.60, MCNPX2.7.0 and Data Libraries.
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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 correlatedparticle and the subsequent interactions as close as possible to the physical behavior. Initially, MCNPPoliMi, a modification of MCNP4C, was developed. The first version was developed in 20012002 and released in early 2004 to the Radiation Safety Information Computational Center (RSICC). It was developed formore »
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