A data analysis code for MCNP mesh and standard tallies

We describe a computer program, provisionally named Krempe, developed for the analysis and presentation of mesh and other tallies from Monte Carlo radiation transport codes MCNP and MCNPX. Data along a line can be extracted from mesh tallies and the profile plotted. Peaks in the profile can be fitted to a Gaussian. We describe the use of such a fit to obtain a measure of the spatial spreading of a particle beam. Smoothing functions are available to average over noise. Additional fitting functions are available for peaks in an energy spectrum. Tallies as a function of energy or other variables can be plotted in a variety of styles. Multiple tallies can be shown together, each with an optional independent scaling factor. A new tally can be defined from linear combination of two tallies. Data is imported from MCTAL files. A list shows available tallies. A sequence of tallies can be plotted in three dimensions. (authors)