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 National Lab. (LANL), Los Alamos, NM (United States)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 5519826
- Report Number(s):
- LA-8176-MS; TRN: 80-004854
- 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]
Certification of MCNP Version 4A for WHC computer platforms. Revision 7
Certification of MCNP version 4A for WHC computer platforms
Technical Report
·
Thu Nov 01 00:00:00 EST 1979
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OSTI ID:5519826
Certification of MCNP Version 4A for WHC computer platforms. Revision 7
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Wed May 03 00:00:00 EDT 1995
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Tue May 07 00:00:00 EDT 1996
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OSTI ID:5519826
Related Subjects
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
NEUTRON TRANSPORT
COMPUTER CODES
PHOTON TRANSPORT
M CODES
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
RADIATION TRANSPORT
654001* - Radiation & Shielding Physics- Radiation Physics
Shielding Calculations & Experiments
654003 - Radiation & Shielding Physics- Neutron Interactions with Matter
NEUTRON TRANSPORT
COMPUTER CODES
PHOTON TRANSPORT
M CODES
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
RADIATION TRANSPORT
654001* - Radiation & Shielding Physics- Radiation Physics
Shielding Calculations & Experiments
654003 - Radiation & Shielding Physics- Neutron Interactions with Matter