Calculation of cell volumes and surface areas in MCNP

Abstract

MCNP is a general Monte Carlo neutron-photon particle transport code which treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces. It is necessary to calculate cell volumes and surface areas so that cell masses, fluxes, and other important information can be determined. The volume/area calculation in MCNP computes cell volumes and surface areas for cells and surfaces rotationally symmetric about any arbitrary axis. 5 figures, 1 table.

Keywords

NUCLEAR PHYSICS AND RADIATION PHYSICS
COMPUTER CODES
M CODES
GAMMA TRANSPORT THEORY
NEUTRON TRANSPORT THEORY
PHOTON TRANSPORT
MONTE CARLO METHOD
NEUTRON FLUX
NEUTRON TRANSPORT
VOLUME
NEUTRAL-PARTICLE TRANSPORT
RADIATION FLUX
RADIATION TRANSPORT
TRANSPORT THEORY
* - RADIATION & SHIELDING PHYSICS- RADIATION PHYSICS
SHIELDING CALCULATIONS & EXPERIMENTS
- RADIATION & SHIELDING PHYSICS- NEUTRON INTERACTIONS WITH MATTER

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