TAILIEUCHUNG - 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

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. | Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the library by using integral data assimilation EPJ Nuclear Sci. Technol. 5 24 2019 Nuclear Sciences A. Rizzo et al. published by EDP Sciences 2019 amp Technologies https epjn 2019056 Available online at https REGULAR ARTICLE Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the library by using integral data assimilation Axel Rizzo1 Claire Vaglio-Gaudard1 Gilles Noguere1 Romain Eschbach1 Gabriele Grassi2 and Julie-Fiona Martin2 1 CEA DEN Cadarache 13108 Saint Paul les Durance France 2 Orano Cycle BU Recyclage Paris France Received 17 April 2019 Received in final form 11 October 2019 Accepted 21 November 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- library slight increases of and 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 .

• Năng lượng
• Comparisons of calculated
• Used nuclear fuels
• Quality of nuclear data
• JEFF 3 1 1 library
• Integral data assimilation
• Summary of Doctoral Thesis in Mathematics
• Differential equations
• Integral equations
• Data assimilation problem
• Evolution equations
• Fluid mechanics
• Generalized Least Square
• Re analyses of MASURCA
• FCA IX criticality experiments
• PROFIL irradiation experiments
• Monte Carlo calculations