Detailed Reactor Analysis Code simplifies Monte-Carlo Transport calculations

A computer code has been developed for quick and accurate nuclear design and analysis studies. The Detailed Reactor Analysis Code (DRAC) uses Monte Carlo Code for Neutron and Photon Transport (MCNP), a well established Monte-Carlo Transport code, to conduct parametric studies of various thermionic space reactor systems. DRAC is an Object-Oriented based program that is preprocessor and a post-processor for MCNP. DRAC allows a user unfamiliar with MCNP to easily construct a complicated input file via a graphical interface and windows environment. DRAC also processes the large output file from MCNP and can display various parameters of interest. Given a desired design goal, DRAC can iterate specified parameters on the design until the goal is achieved without the need for human interaction. Additionally, the program can operate on various computer platforms (i.e. personal computer, Sun workstation, HP (Hewlett Packard) workstation, etc.). 4 refs., 11 figs