Description: 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.
Date: January 1, 1980
Creator: Hendricks, J.S.
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