from Wikipedia, the free encyclopedia
Basic data

developer Los Alamos National Laboratory
Publishing year 1957
Current  version MCNP6.2
operating system platform independent
programming language Fortran 90
category Computer physics , Monte Carlo method , particle transport
License Depending on the version, freely available or proprietary
German speaking No

MCNP , the Monte-Carlo N-Particle Transport Code , is a worldwide widespread reactor physics program for the simulation of nuclear processes. It has been developed by Los Alamos National Laboratory since at least 1957. In the US, it is distributed by the Radiation Safety Information Computational Center in Oak Ridge, Tennessee and internationally by the Nuclear Energy Agency in Paris.

A common version is MCNP5 , the current MCNP6.2 . The MCNP website also contains manuals and release notes as Internet documents, for example Volume I of the MCNP5 manual Overview and Theory .

Brief description

The MCNP program can be used for neutron, photon, electron or coupled neutron, photon and electron transport. It is possible to calculate eigenvalues ​​for critical systems . The code deals with any three-dimensional configuration of materials in geometric cells bounded by planar, second-degree surfaces, or elliptical tori .

Point-by-point cross-sectional data are used. For neutrons, all reactions are taken into account that are specified in a specific evaluation of cross sections (such as ENDF / B-VI). Thermal neutrons are described in both the free gas and the S (α, β) model. For photons, the code takes into account incoherent and coherent scattering, the possibility of fluorescence emission after photoelectric absorption, absorption in pair production with local emission of annihilation and bremsstrahlung . A continuous slowing-down model is used for electron transport, which includes positrons , X-rays and bremsstrahlung , but not external or self-induced fields.

Important standard functions that make MCNP very versatile and easy to use include a powerful general source, critical source and surface source, a rich collection of variance reduction techniques, and an extensive collection of cross section data.

MCNP is one of the type of programs that can fill a physicist's professional life.


The MCNPX program , an abbreviation for Monte Carlo N-Particle eXtended , was initially developed in parallel with the MCNP5 program , also at the Los Alamos National Laboratory, from the 1990s onwards. The first version was released in November 1999. With this program, particle collisions of 34 different types of particles (nucleons and ions) and those of more than 2000 heavy ions can be simulated. The upper limit of the kinetic energy of the particles has also been increased significantly. The focus of the applications should therefore be in accelerator and high-energy physics . The development stages of the program are recorded in documents that are freely accessible.

With version MCNP6 , the programs MCNP5 and MCNPX have been merged and additional options have been added. In contrast to MCNP5 , MCNP6 and MCNPX are subject to certain transfer restrictions of the Department of Energy . Anyone who wants to use the programs must meet specified, not too strict criteria.

application areas

MCNP is mainly used to simulate nuclear processes like nuclear fission , but it is also suitable for simulating particle interactions between neutrons , photons and electrons and other particles. Specific application areas include, for example , nuclear fission and fusion reactor design , nuclear criticality safety, radiation protection , dosimetry , decontamination , radiography , medical physics , detector design and analysis , accelerator construction and decommissioning . The WorldCat bibliographic database contains more than 10,000 works that are dedicated to the MCNP program itself or to applications of the program.

See also

Web links

Individual evidence

  1. Edmond Darrell Cash Well, Cornelius Joseph Everett: A practical manual on the Monte Carlo method for random walk problems . University of California, Los Alamos (New Mexico) 1957 (228 pp., [PDF; accessed June 19, 2018]).
  2. a b A General Monte Carlo N-Particle (MCNP) Transport Code: Monte Carlo Methods, Codes, & Applications Group. Retrieved June 19, 2018 .
  3. a b X-5 Monte Carlo Team: MCNP - A General Monte Carlo N-Particle Transport Code, Version 5: Volume I: Overview and Theory. Retrieved June 19, 2018 .
  4. Michael R. James et al. : MCNPX 2.7.X - New Features Being Developed. 2009, accessed June 21, 2018 .
  5. MCNP - Frequently Asked Questions. Retrieved June 21, 2018 .