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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  235 U,  239 Pu and  241 Pu 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 ( 236 U,  238 Pu,  240 Pu,  242 Pu,  241 Am,  243 Am,  245 Cm) and fission products ( 103 Rh,  153 Eu,  154 Eu) as well as for the  238 U(n,2n) and  237 Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of  144 Ce,  133 Cs,  137 Cs and  106 Ru for the  235 U(n th ,f) and  239 Pu(n th ,f) reactions are also reported.

epj nuclear sciences and technologies

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