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

In order to reach the level of MCNP for the disk source, we started by running the code with a simple point source centered between the two disks (top left). We proceeded to wrap the point source in a cell of graphite analogous to our baked graphite disks (top right). The final step was to actually make the graphite the source.

The plot is of all the steps taken to achieve "real-life" geometry.

1a = tally across the entire disk with a single point source varying the separation.
2a = tally across the entire disk with a point source wrapped in graphite.
3a = tally across the entire disk with graphite as the source.
1b = tally across the far surface with point source.
2b = tally across the far surface with point source wrapped in graphite.
3b = tally across the far surface with graphite as the source.

The same procedure was done for the development of the MCNP code tallying across the near surfaces (point source, point source in graphite, graphite as source).

Graphite Simulation - Far 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.