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Title: Monte Carlo N-Particle Transport Code System Including MCNP6.1.1BETA, MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.

Version 01 MCNP6™ 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. This MCNP6.1.1Beta is a follow-on to the MCNP6.1 production version which itself was the culmination of a multi-year effort to merge the MCNP5™ [X-503] and MCNPX™ [PEL11] codes into a single product. This MCNP6.1.1 beta has been released in order to provide the radiation transport community with the latest feature developments and bug fixes in the code. MCNP6.1.1 has taken input from a group of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Radiation Transport Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5). They have combined their code development efforts to produce this next evolution of MCNP. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keffmore » eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams.« less
Publication Date:
OSTI Identifier:
Report Number(s):
MCNP6.1.1BETA/MCNP(6.1/5/X)-EX; 004594MLTPL00
DOE Contract Number:
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Other Software Info:
Owner Installation: Service de Physique Theorique Contributors: Los Alamos National Laboratory, Los Alamos, New Mexico. MCNP6.1.1Beta includes several significant new capabilities not found in previous versions MCNP and includes the full capabilities of MCNP6.1. • Added Correlated Gamma Multiplicity (CGM) model from LANL • Enabled delayed alphas from nuclear interactions • Allow spontaneous neutron and beta sources • Improved time integration of secondary-particle production from spontaneous decay • Added correlated sampling of Delayed-Particle Production • Several unstructured mesh improvements • Number of digits in the mctal and output files were increased for large integers • Several performance enhancements • Capability to create Cerenkov optical photons from charged particles • Added the Compton Image Tally option • The Cosmic source feature now includes heavy ions • Data updates and bug fixes. KEYWORDS: COMPLEX GEOMETRY; COUPLED; CROSS SECTIONS; ELECTRON; GAMMA-RAY; MONTE CARLO; NEUTRON
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
Contributing Orgs:
Not Specified
Country of Publication:
United States

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