Report on INL Activities for Uncertainty Reduction Analysis of FY11
This report presents the status of activities performed at INL under the ARC Work Package on "Uncertainty Reduction Analyses" that has a main goal the reduction of uncertainties associated with nuclear data on neutronic integral parameters of interest for the design of advanced fast reactors under consideration by the ARC program. First, an analysis of experiments was carried out. For both JOYO (the first Japanese fast reactor) and ZPPR-9 (a large size zero power plutonium fueled experiment performed at ANL-W in Idaho) the performance of ENDF/B-VII.0 is quite satisfying except for the sodium void configurations of ZPPR-9, but for which one has to take into account the approximation of the modeling. In fact, when one uses a more detailed model (calculations performed at ANL in a companion WP) more reasonable results are obtained. A large effort was devoted to the analysis of the irradiation experiments, PROFIL-1 and -2 and TRAPU, performed at the French fast reactor PHENIX. For these experiments a pre-release of the ENDF/B-VII.1 cross section files was also used, in order to provide validation feedback to the CSWEG nuclear data evaluation community. In the PROFIL experiments improvements can be observed for the ENDF/B-VII.1 capture data in 238 Pu, 241 Am, 244 Cm, 97 Mo, 151 Sm, 153 Eu, and for 240 Pu(n,2n). On the other hand, 240,242 Pu, 95 Mo, 133 Cs and 145 Nd capture C/E results are worse. For the major actinides 235 U and especially 239 Pu capture C/E's are underestimated. For fission products, 105,106 Pd, 143,144 Nd and 147,149 Sm are significantly underestimated, while 101 Ru and 151 Sm are overestimated. Other C/E deviations from unity are within the combined experimental and calculated statistical uncertainty. From the TRAPU analysis, the major improvement is in the predicted 243 Cm build-up, presumably due to an improved 242 Cm capture evaluation. The COSMO experiment was also analyzed in order to provide useful feedback on fission cross sections. It was found out that ENDF/B-VII.1 238,240 Pu fission cross sections have improved with respect to VII.0 files while 242 Pu's fission cross section has not.