Monte Carlo N-Particle Transport Code
|Stable release||MCNP 6.1 / August 5, 2013|
|Written in||Fortran 90|
|License||MCNPX Single-User Software License (proprietary)|
||This article needs attention from an expert in Computer science. (February 2011)|
Monte Carlo N-Particle Transport Code (MCNP) is a software package for simulating nuclear processes. It is developed by Los Alamos National Laboratory since at least 1957 with several further major improvements. It is distributed within the United States by the Radiation Safety Information Computational Center in Oak Ridge, TN and internationally by the Nuclear Energy Agency in Paris, France. It is used primarily for the simulation of nuclear processes, such as fission, but has the capability to simulate particle interactions involving neutrons, photons, and electrons. "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."
MCNPX (Monte Carlo N-Particle eXtended) was also developed at Los Alamos National Laboratory, and is capable of simulating particle interactions of 34 different types of particles (nucleons and ions) and 2000+ heavy ions at nearly all energies, including those simulated by MCNP.
MCNP6 is a merger of MCNP5 and MCNPX.
- "MCNP6.1 home Page". LANL. 5 August 2013. Archived from the original on 7 August 2013. Retrieved 7 August 2013.
- Cashwell, E.D.; Everett, C.J. (1959). A Practical Manual on the Monte Carlo Method for Random Walk Problems. London: Pergamon Press.
- Safety code (nuclear reactor)
- Monte Carlo method
- Monte Carlo methods for electron transport
- Nuclear reactor
- Nuclear engineering