MCNP® and Monte Carlo N-Particle® are registered trademarks owned by Los Alamos National Security, LLC, manager and operator of Los Alamos National Laboratory. Any third party use of such registered marks should be properly attributed to Los Alamos National Security, LLC, including the use of the ® designation as appropriate. Please note that trademarks are adjectives and should not be pluralized or used as a noun or a verb in any context for any reason. Any questions regarding licensing, proper use, and/or proper attribution of Los Alamos National Security, LLC marks should be directed to email@example.com.
MCNP6.3: Monte Carlo N–Particle® Transport Code System Version 6.3.
RELATED DATA LIBRARIES
MCNPDATA: Standard Neutron, Photoatomic, Photonuclear, Electron/Positron, Proton, and other Light Ion Data. More information is available at https://nucleardata.lanl.gov
AUXILIARY PROGRAMS included in the distribution
Auxiliary programs are listed in Appendix E of the MCNP6.3 manual included in this distribution package and available at https://mcnp.lanl.gov/manual.htm.
Los Alamos National Laboratory, Los Alamos, New Mexico.
Package ID: C00870MNYCP00:
Fortran 2018, C and C++, Linux, macOS and Windows (C00870MNYCP00).
Package ID: C00870MNYCP01:
Executables only (no source code) for Windows, Linux, and MacOSX systems(C00870MNYCP01)
The MCNP®, Monte Carlo N-Particle® code can be used for general-purpose transport of many particles including neutrons, photons, electrons, ions, and many other elementary particles, up to 1 TeV/nucleon. The transport of these particles is through a three-dimensional representation of materials defined in a constructive solid geometry, bounded by first-, second-, and fourth-degree user-defined surfaces. In addition, external structured and unstructured meshes can be used to define the problem geometry in a hybrid mode by embedding a mesh within a constructive solid geometry cell, providing an alternate path to defining complex geometry. Since the last release of the MCNP® code, major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP6.3 as well as the analysis of results. Details regarding this release are available at https://mcnp.lanl.gov/release_630.html
Tabulated nuclear and atomic data and/or physics models are used to simulate the physics at each collision a particle undergoes during the transport process. Typically, tabulated nuclear and atomic data are used in the low-energy regime for a subset of projectile particles (e.g., neutrons, photons, light ions) and target nuclei. In particular,
To ultimately simulate the particle tracking through the defined geometry, the collision physics interactions, and variance reduction methods, pseudo-random numbers are used to sample the underlying probability density functions that describe each of the event processes. Each history in the simulation uses a unique sequence of pseudo-random numbers and can therefore be considered independent from other histories in the simulation. Throughout the career of each computational particle, various events that occur can be tallied. The MCNP code contains numerous tallies: surface current and flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height tally for particle counts and energy or charge deposition, mesh tallies, radiography tallies, perturbation/sensitivity tallies, and a collection of specialized tally treatments. These tallies and their statistical uncertainties are calculated across the ensemble of independent history tally contributions.
Important standard features that make the MCNP code versatile and easy to use include a powerful general source, criticality source, and surface source; both a fixed-source and k-eigenvalue solution mode; 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. All of the capabilities within the MCNP code can be used on Windows, Linux, and macOS platforms, with the majority of the features capable of parallel execution. The application areas that use the predictions of the MCNP code include (but are not limited to): radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, critical and subcritical experiment design and analysis, detector design and analysis, nuclear oil-well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning, and nuclear safeguards and nonproliferation.
Known issues and limitations with this version are available in the release notes at https://mcnp.lanl.gov/release_630.html with any additional issues identified after the release notes were published identified on the webpage itself.
MCNP6.3 will operate on most computers.
MCNP6.3 runs under Linux, macOS, and Windows. The MCNP6.3.0 distributed executables are built using the Intel compilers. More information on building this version of the MCNP code and other software dependencies is available at https://mcnp.lanl.gov/release_630.html
Included documentation in electronic format to be extracted to your hard drive:
Many more reference documents on MCNP® are included in the MCNP® REFS directory of this distribution.
directory of this distribution.
The MCNP6.3 package is distributed on 1 DVD that can be read on Windows, Linux or macOS systems.
· Disc 1 of C00870MNYCP00 contains MCNP6.3 source code, pre-compiled executables, and reference documents.
· Disc 1 of C00870MNYCP01 contains MCNP6.3 pre-compiled executables and reference documents.
Export control regulations restrict the distribution of source code. Individuals that need the
source code must include a justification as to why the source code is needed that can be
included as a comment when submitting the request for the executable version. FORTRAN
2018, C, C++; Windows PC, Linux PC, macOS
[Package ID: C00870MNYCP00 (full source distribution, justification required) and C00870MNYCP01
12. DATE OF ABSTRACT
KEYWORDS: MONTE CARLO; PARTICLE TRANSPORT; COMPLEX GEOMETRY; NEUTRON, PHOTON; GAMMA-RAY; ELECTRON; ION; COUPLED; CROSS SECTIONS; PHYSICS MODELS