Graphical user input interface for MCNP

The Monte Carlo Neutron Photon (MCNP) computer code is a transport code that includes a powerful three-dimensional geometry and source modeling capability. MCNP has applications in shielding, reactor physics, criticality, and many other areas. In creating an input file, entering laborious descriptions of geometry, sources, tallies, and optimization parameters with a line editor is very tedious and error prone. Input files for MCNP typically contain from hundreds to thousands of lines of data. This makes it difficult during the line-edit sessions to remember exactly what portion of the input file still remains to be completed. Once an input file is created, additional time is required to plot the geometry and correct any errors in the geometry description. To aid in the creation of MCNP input files, a graphical interface has been developed that enables the use to interactively build the geometry with the aid of two or more dynamic cross-sectional views of the model