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Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors
المؤلفون المشاركون
Trivedi, Ishita
Hou, Jason
Grasso, Giacomo
Ivanov, Kostadin
Franceschini, Fausto
المصدر
Science and Technology of Nuclear Installations
العدد
المجلد 2020، العدد 2020 (31 ديسمبر/كانون الأول 2020)، ص ص. 1-14، 14ص.
الناشر
Hindawi Publishing Corporation
تاريخ النشر
2020-08-12
دولة النشر
مصر
عدد الصفحات
14
الملخص EN
In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse.
The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite.
The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients.
These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA.
An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures.
It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections.
The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated.
The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections.
It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively.
Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.
نمط استشهاد جمعية علماء النفس الأمريكية (APA)
Trivedi, Ishita& Hou, Jason& Grasso, Giacomo& Ivanov, Kostadin& Franceschini, Fausto. 2020. Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors. Science and Technology of Nuclear Installations،Vol. 2020, no. 2020, pp.1-14.
https://search.emarefa.net/detail/BIM-1209433
نمط استشهاد الجمعية الأمريكية للغات الحديثة (MLA)
Trivedi, Ishita…[et al.]. Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors. Science and Technology of Nuclear Installations No. 2020 (2020), pp.1-14.
https://search.emarefa.net/detail/BIM-1209433
نمط استشهاد الجمعية الطبية الأمريكية (AMA)
Trivedi, Ishita& Hou, Jason& Grasso, Giacomo& Ivanov, Kostadin& Franceschini, Fausto. Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors. Science and Technology of Nuclear Installations. 2020. Vol. 2020, no. 2020, pp.1-14.
https://search.emarefa.net/detail/BIM-1209433
نوع البيانات
مقالات
لغة النص
الإنجليزية
الملاحظات
Includes bibliographical references
رقم السجل
BIM-1209433
قاعدة معامل التأثير والاستشهادات المرجعية العربي "ارسيف Arcif"
أضخم قاعدة بيانات عربية للاستشهادات المرجعية للمجلات العلمية المحكمة الصادرة في العالم العربي
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