Monte Carlo N-Particle Transport Code

Monte Carlo N-Particle Transport (MCNP) 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 and is developed by Los Alamos National Laboratory. 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.

Point-wise cross section data are typically used, although group-wise data also are available. In MCNP neutron transport calculations, all reactions given in a particular cross-section evaluation (such as the evaluated nuclear data files from ENDF) are accounted for. Thermal neutrons are described by both the free gas and S(α,β) models. MCNP photon transport models account for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption 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.

Features of the MCNP code include a general source, criticality source, and surface source; geometry and output tally plotters; a collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data.

MCNP simulations track the results of its simulation through "tallies". When applied to a surface, a cell or a region of space in general they can track the following quantities: surface current and flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height tally for energy or charge deposition, mesh tallies, and radiography tallies.

The MCNP code is commonly used in situations where direct experimental measurement is impractical, cost-prohibitive, or impossible. Applications include the analysis and design of radiation shielding, nuclear systems, detectors, and radiation sources. The code is distributed with evaluated nuclear data libraries and supports verification and validation activities through benchmark comparisons and regression testing. MCNP's predictive capabilities are considered to be highly reliable by the international community, based on its performance with verification and validation test suites, comparisons to its predecessor codes, automated testing and large amount of work that has been done using it in the past 60 years.

MCNP and the name Monte Carlo N-Particle are registered trademarks of Los Alamos National Laboratory. The software is subject to US nuclear technology export controls