MCNP, a general Monte Carlo code for neutron and photon transport: a summary
The general-purpose Monte Carlo code MCNP can 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).
- Research Organization:
- Los Alamos Scientific Lab., NM (USA)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 5519826
- Report Number(s):
- LA-8176-MS
- Country of Publication:
- United States
- Language:
- English
Similar Records
MCNP: a general Monte Carlo code for neutron and photon transport. [IN Fortran for CDC 7600]
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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
COMPUTER CODES
M CODES
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
NEUTRON TRANSPORT
PHOTON TRANSPORT
RADIATION TRANSPORT
Shielding Calculations & Experiments
654003 -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
COMPUTER CODES
M CODES
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
NEUTRON TRANSPORT
PHOTON TRANSPORT
RADIATION TRANSPORT