Development of a One-Group Cross Section Library for Use in Sensitivity Analyses for Graphite-Moderated Reactors

Several methods have been developed previously for estimating cumulative energy production and plutonium production from graphitemoderated reactors (including the well-known GIRM technique). To facilitate sensitivity analysis of these methods, a one-group cross section and fission product yield library for the fuel and graphite activation products has been developed for MAGNOX-style reactors. This library is intended for use in the ORIGEN computer code which calculates the buildup, decay, and processing of radioactive materials. The library was developed using a fuel cell model in Monteburns. This model consisted of a single fuel rod including natural uranium metal fuel, magnesium oxide (magnox) cladding, carbon dioxide coolant, and Grade A United Kingdom (UK) graphite. The data library has been benchmarked for use with ORIGEN 2.2 and ORIGEN ARP. Using this library a complete sensitivity analysis can be performed for GIRM and other technique. A brief sensitivity analysis is given.