Regular Article

Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the JEFF-3.1.1 library by using integral data assimilation

¹ CEA, DEN Cadarache, 13108 Saint Paul les Durance, France
² Orano Cycle, BU Recyclage, Paris, France

^* e-mail: axel.rizzo@cea.fr

Received: 17 April 2019
Received in final form: 11 October 2019
Accepted: 21 November 2019
Published online: 23 December 2019

Abstract

Comparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations. The experimental database used in the present study − containing more than a thousand isotopic ratio measurements for UOX and MOX fuels with burnup ranging from 10 GWd/t up to 85 GWd/t − allowed to investigate 45 isotopic ratios covering a large number of actinides (U, Np, Pu, Am and Cm) and fission products (Nd, Cs, Sm, Eu, Gd, Ru, Ce, Tc, Mo, Ag and Rh). The Integral Data Assimilation procedure implemented in the CONRAD code was used to provide nuclear data trends with realistic uncertainties for Pressurized Water Reactors (PWRs) applications. Results confirm the quality of the ²³⁵U, ²³⁹Pu and ²⁴¹Pu neutron capture cross sections available in the JEFF-3.1.1 library; slight increases of +1.2 ± 2.4%, +0.5 ± 2.2% and +1.2 ± 4.2% are respectively suggested, these all being within the limits of the quoted uncertainties. Additional trends on the capture cross sections were also obtained for other actinides (²³⁶U, ²³⁸Pu, ²⁴⁰Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, ²⁴⁵Cm) and fission products (¹⁰³Rh, ¹⁵³Eu, ¹⁵⁴Eu) as well as for the ²³⁸U(n,2n) and ²³⁷Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of ¹⁴⁴Ce, ¹³³Cs, ¹³⁷Cs and ¹⁰⁶Ru for the ²³⁵U(n_th,f) and ²³⁹Pu(n_th,f) reactions are also reported.