A consistent evaluation of delayed neutron group constants and covariances for ²³⁵U and ²³⁸U using a combination of microscopic and macroscopic data

Delayed neutron group constants, i.e. decay constants and relative abundances, are important data in reactor kinetics calculations. Due to the current inconsistencies occurring in the time description of delayed neutron emission and the lack of covariance data in ENDF evaluated files, we propose a method to derive a set of eight group relative abundances for the thermal fission of ²³⁵U and fast fission of ²³⁸U, based on a two-step approach. The former considers summation calculations using the JEFF-3.1.1 evaluated fission yields and the ENDF/B-VII.l neutron emission probabilities and half-lives, to evaluate a first set of delayed neutron abundances. Covariances are estimated from a Monte-Carlo approach, sampling the parameters according to their uncertainties. The latter step is to consider two independent measurements of the zero power transfer function using both neutron noise and reactivity modulation techniques. A Bayesian approach was applied, with the microscopic approach used as a prior estimation of the group constants and related covariances and applying the integral data assimilation technique to account for the feedback of the reactor experiments. These data were compared to the evaluated ones from the JEFF-3.3 (Keepin 8-group expanded), ENDF/B-VIII.0β4 (Brady and England) and JENDL-4.0 (Keepin 6-group) libraries, to compute dynamic reactivity and associated uncertainty of various LWR core benchmarks loaded with U fuel, covering ²³⁵U enrichments from 2 to 20%.