Depletion uncertainty analysis to the MYRRHA fuel assembly model

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In this work, the objective is to perform an uncertainty analysis on a MYRRHA -Rev.1.6 irradiation cycle study, being applied to a depletion scenario of a single fresh fuel assembly while assuming reflective boundary conditions. Such analysis is statistically based on the application of Wilk’s method of building tolerance limits after 100 depletion calculations were performed with the SERPENT2 code. Due to the computational burden of such type of simulations, this propagation of nuclear data covariances study (allowed by the fast computational performance of SERPENT2) was done at constant power, constant flux and, in a final exercise, at constant power with the addition of fission yield uncertainties (all of these cases employed ENDF/B-VII.1 data). It was observed that while depleting at constant power, the statistical variation of key fission products such as \textsuperscript{148}Nd is almost not present because of the normalization factor applied to the flux. In contrast, the irradiation at constant flux reveals dependence on burnup. Finally, the added fission yield uncertainties makes clear the fact that they directly impact the degree of final uncertainty computed for fission products exemplified by \textsuperscript{148}Nd and \textsuperscript{135}Xe important for burnup estimation and reactor operation, respectively.


Original languageEnglish
Article number12001
Pages (from-to)1-4
Number of pages4
JournalEPJ Web of Conferences
Publication statusPublished - 30 Sep 2020
Event2019 - ND: International Conference on Nuclear Data for Science and Technology - China National Convention Center, Beijing, China
Duration: 19 May 201922 May 2019


  • Uncertainty, Random sampling, Depletion

ID: 6931893