Depletion uncertainty analysis to the MYRRHA fuel assembly model

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Depletion uncertainty analysis to the MYRRHA fuel assembly model. / Hernandez Solis, Augusto; Stankovskiy, Alexey; Fiorito, Luca; Van den Eynde, Gert.

In: EPJ Web of Conferences, Vol. 239, 12001, 30.09.2020, p. 1-4.

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@article{a6d763310e8c4ad288b8c2fb126fa791,
title = "Depletion uncertainty analysis to the MYRRHA fuel assembly model",
abstract = "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.",
keywords = "Uncertainty, Random sampling, Depletion",
author = "{Hernandez Solis}, Augusto and Alexey Stankovskiy and Luca Fiorito and {Van den Eynde}, Gert",
note = "Score=10",
year = "2020",
month = "9",
day = "30",
doi = "10.1051/epjconf/202023912001",
language = "English",
volume = "239",
pages = "1--4",
journal = "EPJ Web of Conferences",
issn = "2100-014X",
publisher = "EDP Sciences",

}

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TY - JOUR

T1 - Depletion uncertainty analysis to the MYRRHA fuel assembly model

AU - Hernandez Solis, Augusto

AU - Stankovskiy, Alexey

AU - Fiorito, Luca

AU - Van den Eynde, Gert

N1 - Score=10

PY - 2020/9/30

Y1 - 2020/9/30

N2 - 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.

AB - 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.

KW - Uncertainty

KW - Random sampling

KW - Depletion

UR - https://ecm.sckcen.be/OTCS/llisapi.dll/open/40558782

U2 - 10.1051/epjconf/202023912001

DO - 10.1051/epjconf/202023912001

M3 - Article

VL - 239

SP - 1

EP - 4

JO - EPJ Web of Conferences

JF - EPJ Web of Conferences

SN - 2100-014X

M1 - 12001

ER -

ID: 6931893