BISON-C: A One-Dimensional Transport and Burnup Calculation Code with Consideration of Actinides and Fission Products

The BISON-C program for one-dimensional discrete ordinates transport and nuclide production-depletion calculations is described. It works under the UNIX system and allows for fission product transmutation calculations, application of different sets of cross sections for one nuclide in different regions (providing correction for self-shielding effects), as well as for defining a better time-step scheme (improving burnup calculation results). The auxiliary tools for manipulation of the data libraries are also provided. The package is publicly available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory