Abstract
MCNP is a very general Monte Carlo neutron-photon transport code with approximately 250 person years of Group X-6 code development invested. It is highly portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code. Many useful and important variants of MCNP exist for special applications.
The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will most probably include adding a variety of particles to be transported and incorporating adaptive Monte Carlo variance reduction techniques, as well as a number of other physics and user improvements.
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References
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© 1985 Springer-Verlag
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Forster, R.A., Godfrey, T.N.K. (1985). MCNP - a general Monte Carlo code for neutron and photon transport. In: Alcouffe, R., Dautray, R., Forster, A., Ledanois, G., Mercier, B. (eds) Monte-Carlo Methods and Applications in Neutronics, Photonics and Statistical Physics. Lecture Notes in Physics, vol 240. Springer, Berlin, Heidelberg. https://doi.org/10.1007/BFb0049033
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DOI: https://doi.org/10.1007/BFb0049033
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