Validation of MCNP, a comparison with SCALE: Part 1, Highly enriched uranium solutions

Six sets of critical experiments have been modeled and an effective multiplication factor, keff, calculated for each experiment. Each of the experiments used either uranyl nitrate or uranyl fluoride solution as the fissile material. The calculations of keff were performed using MCNP 4. MCNP is a Monte Carlo based transport code which used continuous-energy nuclear data for these calculations. ENDF/B-V cross sections were used for this study. This report also compares the results of MCNP with the results of the CSAS25 module of SCALE4 using the 27 group ENDF/B-IV cross sections. A previous validation study includes information about the CSAS25 module and the resulting data. What follows in Section 2 are descriptions of the experiments performed. Section 3, Results, consists primarily of tables listing the calculated keff's. Lastly, Sections 4 and 5 give a general discussion of observed trends in the results and suggested biases