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Uncertainty Quantification and Propa...

Trivedi, Ishita.

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Uncertainty Quantification and Propagation Methodology for Steady State Safety Analysis of Lead-Cooled Fast Reactors.

Record Type:	Electronic resources : Monograph/item
Title/Author:	Uncertainty Quantification and Propagation Methodology for Steady State Safety Analysis of Lead-Cooled Fast Reactors./
Author:	Trivedi, Ishita.
Published:	Ann Arbor : ProQuest Dissertations & Theses, : 2020,
Description:	69 p.
Notes:	Source: Masters Abstracts International, Volume: 81-11.
Contained By:	Masters Abstracts International81-11.
Subject:	Nuclear engineering. -
Online resource:	http://pqdd.sinica.edu.tw/twdaoapp/servlet/advanced?query=28004254
ISBN:	9798643182320

Uncertainty Quantification and Propagation Methodology for Steady State Safety Analysis of Lead-Cooled Fast Reactors.
Trivedi, Ishita.

Uncertainty Quantification and Propagation Methodology for Steady State Safety Analysis of Lead-Cooled Fast Reactors. - Ann Arbor : ProQuest Dissertations & Theses, 2020 - 69 p.

Source: Masters Abstracts International, Volume: 81-11.

Thesis (M.Sc.)--North Carolina State University, 2020.

This item must not be sold to any third party vendors.

Current efforts towards development of preliminary designs for Lead-cooled Fast Reactors (LFRs) have demonstrated a need for assessing uncertainty propagation through the reactor system. Safety parameters determined using modern codes have a direct impact from nuclear data uncertainties. Evaluating uncertainties will lead to a better understanding of their impact on LFR core design, and will identify the design safety limits that are reviewed during the licensing process. Here, "Best Estimate Plus Uncertainty" approach is applied to LFR to propagate nuclear data uncertainties through multiple scales of core modelling. The Monte Carlo code SERPENT-2.0 with implemented Generalized Perturbation Theory (GPT) was used to calculate the sensitivity coefficients of the multiplication factor with respect to nuclide and reaction-dependent nuclear data for fuel assembly models on fuel lattice level. Nuclear data UQ&P was then extended to whole core using Argonne National Lab (ANL) Advanced Reactor Computational (ARC) suite and verified with SERPENT-2.0. In ARC, DIF3D was employed for core modeling and PERSENT was used for sensitivity coefficient calculations. Standard deviations for reactivity feedback coefficients such as doppler, radial expansion and fuel/structure/cooland density were determined. From initial assessment, main contributors of uncertainty were traced back to a number of common nuclide reaction pairs including U-235 capture, U-238 capture, U-238 elastic, U-238 inelastic, and Pu-239 capture.

ISBN: 9798643182320Subjects--Topical Terms:

595435
Nuclear engineering.
Subjects--Index Terms:

Lead-cooled fast reactors(LFRs)

Uncertainty Quantification and Propagation Methodology for Steady State Safety Analysis of Lead-Cooled Fast Reactors.
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520 $a Current efforts towards development of preliminary designs for Lead-cooled Fast Reactors (LFRs) have demonstrated a need for assessing uncertainty propagation through the reactor system. Safety parameters determined using modern codes have a direct impact from nuclear data uncertainties. Evaluating uncertainties will lead to a better understanding of their impact on LFR core design, and will identify the design safety limits that are reviewed during the licensing process. Here, "Best Estimate Plus Uncertainty" approach is applied to LFR to propagate nuclear data uncertainties through multiple scales of core modelling. The Monte Carlo code SERPENT-2.0 with implemented Generalized Perturbation Theory (GPT) was used to calculate the sensitivity coefficients of the multiplication factor with respect to nuclide and reaction-dependent nuclear data for fuel assembly models on fuel lattice level. Nuclear data UQ&P was then extended to whole core using Argonne National Lab (ANL) Advanced Reactor Computational (ARC) suite and verified with SERPENT-2.0. In ARC, DIF3D was employed for core modeling and PERSENT was used for sensitivity coefficient calculations. Standard deviations for reactivity feedback coefficients such as doppler, radial expansion and fuel/structure/cooland density were determined. From initial assessment, main contributors of uncertainty were traced back to a number of common nuclide reaction pairs including U-235 capture, U-238 capture, U-238 elastic, U-238 inelastic, and Pu-239 capture.
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