MCNP - Monte Carlo N-Particle

MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron 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 elliptical tori. Point wise 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(alpha,beta) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields. Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data.

Graphite Simulation - Near Side

We put a graphite disk source between two .01 inch disks that have the same radius as our baked graphite disks.Then, the two surfaces were put 1 cm apart with the disk source centered between them. Tallying .511 MeV photons across the total disk area and the near surfaces of the disks, we ran MCNP from a 1 cm to 4 cm separation with a .5 cm step.The plot is tallies per particle of the near surface of the disk only as a function of seperation.Analyization of the data is still taking place to help explain the unexpected results.