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CCC-0740 MCNP5/MCNPX.

MCNP5/MCNPX, Monte Carlo N-Particle Transport Code System Including MCNP5 1.50, MCNPX 2.6.0, VISED and Data Libraries

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1. NAME OR DESIGNATION:  ******************************************************************************
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MCNP5/MCNPX. This package contains MCNP5 1.50, MCNPX 2.6.0, VISED 22S, VISED X_22S and MCNPDATA.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.

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Machines used:

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3. DESCRIPTION OF PROGRAM OR FUNCTION

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MCNP5 1.50
==========
MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for critical systems. Some of the new features of MCNP5 1.50 include:
- Variance reduction with pulse height tallies
- New VAR input card added to control variance reduction methods
- Annihilation gamma tracking
- Doppler broadening added to the makxsf utility code
- Improved S(alpha,beta)thermal scattering
- Large lattice enhancements
- Direct RSSA file reading for distributed multiprocessing
- Improve Compton scattering PSC calculation for detectors & DXTRAN
- Web-based documentation
  
See the MCNP home page for up-to-date information http://mcnp-green.lanl.gov/ with a link to the MCNP Forum. For information on user experiences with MCNP, see the Electronic Notebook at http://rsicc.ornl.gov. Additional information posted by MCNP developers to the MCNP Forum can be viewed on the web in the MCNP5 electronic notebook (http://www-rsicc.ornl.gov/rsiccnew/CFDOCS/mcnp5enotebook.cfm) in which the Forum entries are archived.
  
MCNPX 2.6.0
===========
MCNPX (MCNP eXtended) is a Fortran90 (F90) Monte Carlo radiation transport computer code that transports many particles over a broad range of energies. It is a superset of MCNP4C3 and has many capabilities beyond MCNP4C3. MCNPX is a production computer code that models the interaction of radiation with matter. New capabilities and enhancements of MCNPX 2.6.0 beyond MCNPX 2.5.0 are listed below. For details, see LA-UR-08-2216.pdf posted on the MCNPX website http://mcnpx.lanl.gov/.
- Depletion/burnup;
- Heavy-ion (Z>2) transport;
- LAQGSM physics model;
- CEM03 physics;
- Long file names;
- Delayed-particle emission;
- Criticality source convergence acceleration;
- Energy-time weight windows;
- Spherical mesh weight windows;
- Charged ions from neutron capture in table range;
- Tallies terminated at desired precision: STOP card;
- Numerous corrections/enhancements/extensions;
- Muon capture physics.
  
MCNPX is a general purpose Monte Carlo radiation transport code that tracks nearly all particles at nearly all energies. The official release date of MCNPX 2.6.0 is April 30, 2008. MCNPX began in 1994 as a code-merger project of MCNP 4B and LAHET 2.8. It was first released to the public in 1999 as version 2.1.5. In 2002, MCNPX was upgraded to MCNP 4C, converted to Fortran 90, enhanced with 12 new features, and released to the public as version 2.4.0. The release of version 2.6.0 includes many new features described in the "MCNPX Extensions Version 2.6.0" document which is provided with the MCNPX distribution. The depletion/burnup capability is based on CINDER90 and MonteBurns. MCNPX depletion is a linked process involving steady-state flux calculations in MCNPX and nuclide depletion calculations in CINDER90. Currently, the depletion/burnup/transmutation capability is limited to criticality (KCODE) problems. Physics improvements include a new version of the Cascade-Exciton Model (CEM), the addition of the Los Alamos Quark-Gluon String Model (LAQGSM) event generator, and a substantial upgrade to muon physics. Current physics modules include the Bertini and ISABEL models taken from the LAHET Code System (LCS), CEM 03, and INCL4. Many new tally source and variance-reduction options have been developed. MCNPX is released with libraries for neutrons, photons, electrons, protons and photonuclear interactions. In addition, variance reduction schemes (such as secondary particle biasing), and new tallies have been created specific to the intermediate and high energy physics ranges. The 'mesh' and 'radiography' tallies were included for 2 and 3-dimensional imaging purposes. Energy deposition received a substantial reworking based on the demands of charged-particle high-energy physics. An auxiliary program, GRIDCONV, converts the mesh and radiography tally as well as standard MCTAL-file results for viewing by independent graphics packages. The code may be run in parallel at all energies via PVM or MPI. Information about MCNPX development can be found on the web site http://mcnpx.lanl.gov/.
  
MCNPDATA
========
These cross-section libraries are released for use with the MCNP/MCNPX Monte Carlo code packages. This release includes all of the LANL X-Division distributed neutron and proton data libraries, the photoatomic libraries, photonuclear data library LA150U, the electron libraries EL1 and EL03, an updated XSDIR file, and a SPECS file for use with MAKXSF to convert the ascii data libraries into binary form. This release is intended to completely replace previous RSICC releases ZZ-MCNPDAT (DLC-0105), ZZ-MCNPDAT6 (DLC-0181), ZZ-MCNPXS (DLC-0189), ZZ-MCNPDATA (DLC-0200), ZZ-MCNPXDATA (DLC-0205) as well as the cross sections previously included with CCC-0200/MCNP4A, CCC-0710/MCNP5 1.30, and CCC-0730/MCNP5 1.40/MCNPX 2.5.0. The release will be updated as new libraries become available.
  
This readme_data.txt file provides information regarding the data libraries contained in this release. The XSDIR file is specific to this release and may not work with previous packages. As of the November 2007 release (with MCNP5_RSICC_1.50), the default continuous energy neutron transport data with 389 isotopes and 3 elements are from the ENDF70 library (based upon the ENDF/B-VII.0 evaluations).  The default proton transport data are from the ENDF70PROT library.  Older data can be accessed by specifically requesting the identifier associated with each library.
  
The libraries MCPLIB04 and EL03 are the default libraries for photoatomic and electron transport respectively. ENDF70SAB is the default library for S(a,b), and LA150U is the default library for photonuclear transport.  More information on the data libraries contained in this release is available in Appendix G of the MCNP5 manual or from the data team's web site at http://www-xdiv.lanl.gov/PROJECTS/DATA/nuclear/
  
Continuous-energy Neutron Data Libraries:
The following lists the publicly released continuous-energy neutron data libraries for use with MCNP as of May 2008.
- endf70(a-k) - ENDF/B-VII Release 0
- t16_2003 - pre ENDF/B-VII.0 evaluations from Los Alamos Group T-16 for 15 isotopes
- actia and actib - ENDF/B-VI Release 8
- endf66(a-e) - ENDF/B-VI Release 6
- la150n: 150-MeV Neutron Library for MCNP
- uresa - ENDF/B-VI Release 4 with probability tables
- endf6dn - ENDF/B-VI Release 2 with delayed-neutron data
- endf62mt - multitemperature ENDF/B-VI Release 2
- endf60 - ENDF/B-VI Release 2
- newxs - LANL based evaluations
- rmccs - ENDF/B-V and LANL based evaluations
- rmccsa - ENDF/B-V and LANL based evaluations
- endf5p - ENDFB-V
- endf5u - ENDF/B-V
- misc5xs - Contains a number of previously released small libraries
- kidman - fission product evaluations
- 100xs - LANL based evaluations for a subset of isotopes up to 100 MeV
- endl92 - 1992 ENDL library from Lawrence Livermore National Lab (LLNL)
- endf5mt - Multitemperature data previously released as eprixs and u600k
  
Discrete Neutron Libraries:
Discrete neutron cross sections are used with the DRXS input card in MCNP. The currently supported discrete neutron libraries are:
- newxsd - discrete version of newxs
- drmccs - discrete version of rmccs and rmccsa
- dre5 - discrete version of endf5u and endf5p
  
Photoatomic Data Libraries: There are now four photoatomic transport data libraries; mcplib, mcplib02, mcplib03, and mcplib04.
  
MCNP Multigroup Data Libraries: The multigroup neutron and photon library mgxsnp is provided for use with MCNP, and is based primarily on ENDF/B-V evaluations.
  
Photonuclear Data Libraries: one photonuclear library, la150u currently supported.
  
Thermal Neutron Data Libraries:
S(alpha,beta) data are contained in the tmccs, therxs, sab2002, and sab2007, endf70sab libraries. The endf70sab libraries are the default as of the May 2008 release.
  
Electron Data Libraries: el and el03 are the electron transport libraries.
  
Dosimetry Data Libraries: 531dos, 532dos, and llldos are the publicly released dosimetry data libraries.
  
Proton Data Libraries: endf70prot and la150h are the proton transport libraries.
  
The data libraries, as distributed, are in ASCII, or type 1, format.
  
For available documentation on each library, see http://www-xdiv.lanl.gov/projects/data/nuclear/mcnpdata.html
  
MCNPX ascii data libraries
==========================
  
In addition to the standard data libraries for MCNP and MCNPX, some ascii data libraries specifically for use with MCNPX are included in this distribution.
  
Atab.dat ----- data required for LAQGSM model / heavy ions
barpol.dat ----- data required for all MCNPX
bcdlib ----- data for PHTLIB to build binary PHTLIB data file
bcdtp ----- data for BERTIN to build binary BERTIN data file
channel1.tab ----- data required for LAQGSM model / heavy ions
cinder.dat ----- data required for depletion/burnup
cindergl.dat ----- data required for delayed gamma lines
flalpha.tab ----- data required for INCL4 physics
frldm.tab ----- data required for INCL4 physics
gamman.tbl ----- data required for CEM03 physics
gdr.dat ----- data required for photonuclear models
level.tbl ----- data required for CEM03 physics
pace2.data ----- data required for INCL4 physics
shell.tbl ----- data required for CEM03 physics
vgsld.tab ----- data required for INCL4 physics
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4. METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODSE

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODSE

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.METHODS

MCNP5 1.50
==========
The MCNP5 code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including K x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. Electron transport does not include the effects of external or self-induced electromagnetic fields. Photonuclear physics is available for a limited number of isotopes.
  
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data. Energy ranges are from 10-11 to 20 MeV for neutrons with data up to 150 MeV for some nuclides, 1 keV to 1 GeV for electrons, and 1 keV to 100 GeV for photons.
  
MCNPX 2.6.0
===========
All capabilities of MCNP4C3 have been retained. Consult the MCNPX User's Manual for applicability to high energy applications.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

MCNP5 1.50: Energy ranges are noted above.
  
MCNPX 2.6.0
===========
All standard MCNP neutron libraries over their stated ranges (~0-20 MeV).
Neutrons in the LA150 library from 0.0 - 150.0 MeV in tabular range for 42 isotopes (except for 9Be to 100 MeV).
Neutrons from 1.0 MeV in the physics model regime.
Photons from 1 keV - 100 GeV.
Photonuclear interactions from 1.0 to 150.0 MeV in tabular range for 12 isotopes.
Photonuclear interactions from 1.0 MeV in the CEM physics model.
Electrons from 1 keV - 1 GeV.
Protons from 1.0 to 150.0 MeV in tabular range for 41 isotopes.
Protons from 1.0 MeV in the physics model regime.
Pions, muons, and kaons are treated only by physics models.
Light ions from 1 MeV/nucleon in the physics model regime.
Heavy ions from 3 MeV/nucleon in the LAQGSM physics model.
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6. TYPICAL RUNNING TIME

MCNP5 1.50: On a 2 GHz Pentium 4, compilation of MCNP5 takes about 3 minutes, and the 50 regression test cases run in about 2 minutes.
  
MCNPX 2.6.0: Runtimes vary depending on computer speed and problem parameters. On a 2 GHz Pentium 4, compilation of MCNPX takes about 5 minutes. Test cases run in about 5 minutes.
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8. RELATED OR AUXILIARY PROGRAMS

RELATED DATA LIBRARY:
MCNPDATA: Standard Neutron, Photoatomic, Photonuclear, and Electron Data Libraries for MCNP5 and MCNPX.
  
Documentation on the data libraries may be found in Appendix G of the MCNP5 manual (Volume I) and on the web http://www-xdiv.lanl.gov/projects/data/nuclear/mcnpdata.html.
A separate test library of cross sections is used for running regression sample problems, but the test library is not suitable for real problems.
  
MCNP5 1.50
==========
AUXILIARY PROGRAMS included in the distribution:
MAKXSF: Prepares MCNP cross-section libraries; now with Doppler broadening.
ONEGXS: Create 1-group cross-sections with P0 or P1 scattering in ACE format.
VISED 22S: Visual Editor for interactively constructing & visualizing MCNP geometry (runs on Windows PCs only). More VISED information is on the web at the Visual Consultants website http://www.mcnpvised.com/.
  
MCNPX 2.6.0
===========
RELATED DATA LIBRARIES:
MCNPDATA: Standard Neutron, Photoatomic, Photonuclear, and Electron Data Libraries for MCNP5 and MCNPX.
  
AUXILIARY PROGRAMS included in the distribution:
GRIDCONV: Converts output of mesh and radiography tallies to input for external graphics programs.
HTAPE3X: Postprocessor for MCNPX HISTP output.
MAKXSF: Prepares MCNPX Cross-Section Libraries.
HCNV and TRX: Convert LAHET ASCII data to binary.
XSEX3: Analyzes a HISTP history file and generates double-differential particle production cross sections for primary beam interactions
VISED X_22S: Visual Editor for interactively constructing & visualizing MCNPX geometry (runs on Windows PCs only). More VISED information is on the web at the Visual Consultants website http://www.mcnpvised.com/.
  
Running MCNPX requires the included MCNPDATA continuous energy cross-section data. This library distribution includes the standard MCNP libraries, along with the LA150H proton data tables for 41 isotopes and the ASCII data files mentioned above. It should be noted that only the LA150 libraries (LA150N, LA150U, LA150H) contain emission data for light ions (i.e., Z=1-2), and in many cases only above 20 MeV (details are provided in the MCNPX User's Manual).
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9. STATUS

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10. REFERENCES

MCNPX 2.6.0:
Background references:
- D. B. Pelowitz, ed., "MCNPX User's Manual, Version 2.5.0," LA-CP-05-0369 (April 2005).
- L. S. Waters, Ed., "MCNPX User's Manual, Version 2.4.0," LA-CP-02-408 (Sept. 2002).
- L. S. Waters, Ed., "MCNPX User's Manual, Version 2.3.0," LA-UR-02-2607 (April 2002).
- J. F. Briesmeister, Ed., "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C," LA-13709-M (April 2000).
- R. E. Prael and H.Lichtenstein, "User Guide to LCS: The LAHET Code System," LA-UR-89-3014, Revised (September 15, 1989).
- M. B. Chadwick, P. G. Young, S. Chiba, S. C. Frankle, G. M. Hale, H. G. Hughes, A. J. Koning, R. C. Little, R. E. MacFarlane, R. E. Prael, and L. S. Waters, "Cross Section Evaluations to 150 MeV for Accelerator-Driven Systems and Implementation in MCNPX," Nuclear Science and Engineering 131, Number 3 (March 1999) 293.
- M. B. Chadwick, P. G. Young, R. E. MacFarlane, P. Moller, G. M. Hale, R. C. Little, A. J. Koning and S. Chiba, "LA150 Documentation of Cross Sections, Heating, and Damage: Part A (Incident Neutrons) and Part B (Incident Protons)," LA-UR-99-1222 (1999).
- S. G. Mashnik, A. J. Sierk, O. Bersillon, and T. A. Gabriel, "Cascade- Exciton Model Detailed Analysis of Proton Spallation at Energies from 10 MeV to 5 GeV," Nucl. Instr. Meth. A414 (1998) 68. (Los Alamos National Laboratory Report LA-UR-97-2905).
- Stepan G. Mashnik, Konstantin K. Gudima, Arnold J. Sierk, Mircea I. Baznat, and Nikolai V. Mokhov, "CEM03.01 User Manual," Los Alamos National Laboratory report LA-UR-05-7321 (2005); RSICC Code Package PSR-532, http://www-rsicc.ornl.gov/codes/psr/psr5/psr-532.html/ (2006).
- S. G. Mashnik, K. K. Gudima, M. I. Baznat, A. J. Sierk, R. E. Prael, and N. V. Mokhov, "CEM03.S1, CEM03.G1, LAQGSM03.S1, and LAQGSM03.G1 Versions of CEM03.01 and LAQGSM03.01 Event-Generators," Los Alamos National Laboratory report LA-UR-06-1764 (March 6, 2006), also available at http://mcnpx.lanl.gov > documents.
- W. B. Wilson et al., "Recent Development of the CINDER'90 Transmutation Code and Data Library for Actinide Transmutation Studies," Proc. GLOBAL'95 Int. Conf. on Evaluation of Emerging Nuclear FuelCycle Systems, Versailles, France, p. 848, September 11-14 (1995).
  
MCNPDATA: For available documentation on each library, see http://www-xdiv.lanl.gov/projects/data/nuclear/mcnpdata.html
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11. HARDWARE REQUIREMENTS

MCNP5 is operable on PC's and on Macintosh (Intel-based and PowerPC-based) computers. Expanding and compiling the code system requires 500 MB, and the ASCII cross sections require ~11 GB of hard disk space. Up to 14 GB may be required during file expansion.
  
MCNPX runs under Unix, Linux, and Windows operating systems and has been implemented on various 32-bit and 64-bit workstations and personal computers.  The compiled version of the code tends to run ~8 Mbytes. Dynamic allocation makes memory demands variable on all platforms.
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12. PROGRAMMING LANGUAGE(S) USED

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13. SOFTWARE REQUIREMENTS

MCNP 1.50
=========
MCNP5 1.50 runs on PC's under Windows or Linux and on Macintosh (Intel-based and PowerPC-based) computers with MacOSX. The code has improved support for newer Fortran 90 compilers. The gcc C compiler is used for all versions. The compilers supported are listed below.
- Linux (32-bit and 64-bit): Absoft 9.0 mpi; g95 (0.9) mpi; gfortran (4.4.2) mpi; Intel (9.0, 9.1, 10.0) omp, mpi; Pathscale (3.0) omp, mpi; Portland (6.0, 6.1, 6.2, 7.0) omp, mpi
- Windows: Absoft (9, 10)** mpi; Compaq (6.6B) mpi; g95 (.91)**; gfortran (4.2.1)**; intel (9.1, 10.0) omp, mpi; lahey mpi; portland group (7.1) omp
** = Plotting is not supported with these compilers.
- Macintosh - Intel-based: Intel 10.1 mpi, omp; Absoft 10.1 mpi; g95 (.91) mpi
- Macintosh - PowerPC-based: IBM XLF 8.1 mpi; Absoft 9.2 mpi; g95 (.91) mpi
  
Included executables and reference templates for output and tally files used in the MCNP5 regression test suite were generated for Linux and Windows with the following compilers:
- Windows: 32-Bit Pentium IV, Windows 2000 Professional SP4 with cygwin-1.5.24-2, Intel 10.0.27 f90, gcc 4.2.1, MPICH2 1.0.3-1.
- Linux: 64-bit AMD Opteron 2.0 GHz, Red Hat Linux 9, Intel 10.0.023 f90, gcc. Also included are MCNP5 Linux 32-bit executables - with and without OMP.
- Mac: Intel-based and PowerPC-based executables are included.
  
  Compilation of the MCNP5 source code requires both Fortran 90 and ANSI C standard compilers. Dynamic storage allocation is available on all supported systems. Additional information may be posted to the MCNP Forum archives in the electronic notebook on the RSICC website.
  The 'omp' notation after a compiler indicates that threading (also called "shared memory multiprocessing" or microtasking) can be used when building an MCNP5 executable. For computers having multi-core processors (e.g., Intel Core2 duo, Intel quad-core Xeon, etc.), MCNP5 should generally be compiled with the 'omp' option. Note that some compilers are not yet suitable for compiling a threaded version of MCNP5.
  For compiling MCNP5 on Windows PCs, the Cygwin environment must first be installed. Cygwin is a collection of GNU-based Unix utilities which have been ported to Windows. The Cygwin environment may be obtained at no cost from the web site http://www.cygwin.com.
  For plotting geometry, cross-sections, or results, X11 must be installed on your PC. An X-windows server is required to display the X11 graphics. Suggested servers include ReflectionX, Exceed, and XFree86.
  
MCNPX 2.6.0
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C and Fortran 90 compilers are required to compile. The GNU make utility is required to build the system on Unix and Linux platforms. The GNU make.exe utility is included for Windows users.  The only graphics support for this release is X11 http://www.x.org/. This is a Fortran 90 version of MCNPX which uses standard F90 allocation schemes for dynamic variables on all platforms. The package includes MCNPX 2.6.0 executables created by the LANL developers for the systems listed below. Each of these files contains precompiled executables and corresponding binary data libraries for bertin and phtlib.
  
- Win32.zip: Windows executables, with Intel 9.1 on a 32-bit XP OS.
- Win32I8.zip: Windows 8-byte integer executables, with Intel 9.1 on a 32-bit XP OS.
- Win32MPI.zip: Windows MPICH-2 executables, with Intel 9.1 on a 32-bit XP OS.
- Win32MPII8.zip: Windows 8-byte integer MPICH-2 executables, with Intel 9.1 on a 32-bit XP OS.
- Win32CVF.zip: Windows executables, with CVF 6.6 on a 32-bit XP OS.
- Linux.tar.gz: Linux executables, with INTEL 9.1 on a 64-bit GNU/Linux OS.
- SUN.tar.gz: Sun executables, with XXX on a 32-bit Solaris OS.
  
RSICC tested this release on the following systems:
- AMD Opteron running RedHat Enterprise Linux 4 with Intel 9.1.
- Pentium 4 running Windows XP with Intel 9.1
- DEC6600 running Tru64 Unix V5.1A with HP Fortran V5.5A and C V6.4.
  
WinZIP 8.0 is required to expand mcnpx files under Windows.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

An optional Windows installer is included to install MCNPDATA. The installer uses the open-source 7z.exe program to expand the files under Windows then sets environmental variable DATAPATH.
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by:
                Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.

Developed by:   Los Alamos National Laboratory,
                Los Alamos, New Mexico, U.S.A.
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16. MATERIAL AVAILABLE
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17. CATEGORIES
  • C. Static Design Studies
  • J. Gamma Heating and Shield Design

Keywords: Monte Carlo method, charged particles, complex geometry, coupled neutron gamma cross sections, criticality, electrons, gamma ray, high energy, kaon, neutron, pion, protons, radiation transport, radiography, shielding, spallation, variance reduction.