Investigation of the "summation" method for predicting group dependent delayed neutron data

Abstract

The objectives of this thesis were 1) to propose modifications of the ENDF/BVI delayed neutron data, 2) to derive new delayed neutron parameters for groups 1, 2, 3, and 4 using the summation method, and 3) to establish the relative importance of individual parameters in delayed neutron data sets. The ENDF/B-VI delayed neutron parameters for groups 1 and 2 were reevaluated using a different database than the one used originally. The modified set of delayed neutron data decreases the value of the average delayed neutron lifetime by 15% for U-235 thermal fission, which agrees more closely with Keepin's results. The modified delayed neutron data set also improves the calculated reactor period (relative to the reactor period calculated using Keepin's delayed neutron data). In the new six group set of delayed neutron data, the decay constant of group 1 (Al) is equal to the decay constant of the delayed neutron precursor Br-87 for any fissioning system. A second set of delayed neutron parameters (for groups 1, 2, 3, and 4) were also obtained using the summation method and more recent database. The reactor period calculated using this set agrees within 7% of the reactor period calculated using Keepin's delayed neutron parameters. Finally, the relative importance of each of the thirteen delayed neutron parameters was established. The calculated reactor period is found to be most sensitive to the value of the delayed neutron fraction (fl), the decay constants for groups 1, 2, and 4, (i.e., Al, A2, and A4), and the relative abundances for groups 2, 3, and 4, (i.e., a2, a3, and a4). On the other hand, groups 5 and 6 (the shortest lived groups) are shown to have relatively little effect on the calculated reactor period, especially in the negative reactivity range.