MCNP - A General Monte Carlo N-Particle Transport Code - Version 5


MCNP Users:
The MCNP team would like to notify users of an update to the RSICC release of MCNP from MCNP 5.1.30 to MCNP 5.1.40. No patch file is provided to update previous versions of MCNP. Users are recommended to obtain the latest version directly from RSICC (http://www-rsicc.ornl.gov/).

MCNP 5.1.40 has seven new features, two minor features, minor code improvements, and improved compiler/platform support. A complete description along with a list of bug fixes are listed in the release notes.
Release Notes (in pdf format)

The new MCNP/MCNPX and Data package is distributed on a DVD.

Diagnostics Applications Group - Los Alamos National Laboratory

ABSTRACT

MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, Detector Design and analysis, nuclear oil well logging, Accelerator target design, Fission and fusion reactor design, decontamination and decommissioning. The 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 typically are used, although group-wise data also are available. 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, absorbtion in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields.

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.

MCNP5 contains numerous flexible tallies: surface current & flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height taly for energy or charge deposition, mesh tallies, and radiography tallies.