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Journal Articles

Uncertainty quantification of $$^{237}$$Np, $$^{241}$$Am, and $$^{243}$$Am reaction rates in highly enriched uranium fuel cores at Kyoto University Critical Assembly

Pyeon, C. H.*; Oizumi, Akito; Katano, Ryota; Fukushima, Masahiro

Nuclear Science and Engineering, 199(3), p.429 - 444, 2025/03

https://doi.org/10.1080/00295639.2024.2380624
 Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)

Experimental analyses of neptunium-237 ($$^{237}$$Np), americium-241 ($$^{241}$$Am), and $$^{243}$$Am fission and $$^{237}$$Np capture reaction rates are conducted by the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5, using experimental data at neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly-enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of fission and capture reaction rate ratios of test samples of $$^{237}$$Np, $$^{241}$$Am and $$^{243}$$Am with reference samples of uranium-235 ($$^{235}$$U) and gold-197 ($$^{197}$$Au) are evaluated by the MARBLE code system. In terms of fission reaction rate ratios of $$^{237}$$Np/$$^{235}$$U, $$^{241}$$Am/$$^{235}$$U and $$^{243}$$Am/$$^{235}$$U, a comparison between experiments and Serpent 2 calculations shows an accuracy about 5, 15 and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For capture reaction rate ratios of $$^{237}$$Np/$$^{197}$$Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of $$^{237}$$Np/$$^{235}$$U, $$^{241}$$Am/$$^{235}$$U and $$^{243}$$Am/$$^{235}$$U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% of $$^{243}$$Am/$$^{235}$$U with ENDF/B-VIII.0 at the intermediate neutron spectrum.

Oral presentation

Preliminary study of multi-group cross-section perturbation on random-sampling-based uncertainty analysis

Fujita, Tatsuya

no journal, , 

Past studies have mentioned that the treatment of implicit effect for cross-section perturbation affects the sensitivity coefficients and then the uncertainty analysis results for the k-infinity. The several approaches to consider the above implicit effect has also been discussed for the random-sampling-based uncertainty analysis. In this study, the influence due to implicit effect on typical nuclides and nuclear reactions in PWR 17$$times$$17 UO$$_{2}$$ and MOX fuel assemblies was confirmed prior to treat the implicit effect on the random-sampling-based uncertainty analysis in future studies. In the UO$$_{2}$$ fuel assembly, the influence due to implicit effect on the k-infinity was small, thus the uncertainty quantification only considering the explicit effect would be applicable. On the other hand, further discussion about the influence due to implicit effect was necessary for the MOX fuel assembly.

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