AcknowledgementsAuthors wish to thank the member of the nuclear data processing working group of the JENDL committee for their expert advices and encouragements. Authors also wish to thank JAEA colleagues for their expert advices and suggestions. They helped to understand the methods using the current nuclear data processing systems and their expert advices and suggestions considerably contributed to develop FRENDY.JAEA developed the new nuclear data processing system FRENDY in order to solve the problems of the current widely used nuclear data processing systems and process the new evaluated nuclear data file. Verification of FRENDY was carried out by three steps, i.e., verification of each function, comparison of the results, and comparison of the k eff values for the 79 benchmark experiments in the ICSBEP handbook using cross section data library processed by FRENDY with those by NJOY99. These results verified that FRENDY generates the ACE file correctly.
Burnup is an important parameter in criticality safety evaluations of spent nuclear fuel in which burnup credit is taken into account. The Neodymium-148 method is widely used to evaluate the burnup of post irradiation examination (PIE) samples, and it is well known for its good accuracy. However, accuracy of the evaluated burnup values may be affected by the neutron capture reaction of 147 Nd and 148 Nd. Moreover, in the analysis of PIE data from a PWR, the calculation results of 148 Nd have more than a 1% deviation from the experiment. In this study, the contribution of neutron capture reactions of 147 Nd and 148 Nd to the amount of 148 Nd is discussed. The PIE data analyses using new evaluation of 147 Nd capture cross section show that the JENDL-3.2 cross section data is overestimated. The change in the amount of 148 Nd due to both reactions is less than 0.7% under normal reactor operation conditions. In particular, it is in the 0.1% range if burnup is approximately 30 GWd/t for a BWR and 40 GWd/t for a PWR.
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