MCNP: a general Monte Carlo code for neutron and photon transport. [IN Fortran for CDC 7600]

Technical Report · Thu Nov 01 04:00:00 EST 1979

DOI:https://doi.org/10.2172/5823239· OSTI ID:5823239

The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(..cap alpha..,..beta..) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables.

• Research Organization: Los Alamos Scientific Lab., NM (USA)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 5823239
• Report Number(s): LA-7396-M(Rev.)
• Country of Publication: United States
• Language: English

Related Subjects

654001* -- Radiation & Shielding Physics-- Radiation Physics
Shielding Calculations & Experiments
654003 -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
CDC COMPUTERS
COMPUTER CODES
COMPUTERS
CRAY COMPUTERS
DATA COVARIANCES
EIGENVALUES
FORTRAN
GAMMA TRANSPORT THEORY
M CODES
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
NEUTRON TRANSPORT
NEUTRON TRANSPORT THEORY
PHOTON TRANSPORT
PROGRAMMING LANGUAGES
RADIATION TRANSPORT
THREE-DIMENSIONAL CALCULATIONS
TRANSPORT THEORY