Bootstrapped-ensemble-based Sensitivity Analysis of a trace thermal-hydraulic model based on a limited number of PWR large break loca simulations

Maio, Francesco Di; Bandini, Alessandro; Zio, Enrico; Alberola, Sofía Carlos; Sánchez-Saez, Francisco; Martorell, Sebastián

doi:10.1016/j.ress.2016.04.013

Cited by 14 publications

(21 citation statements)

References 40 publications

(29 reference statements)

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“…Others propose the use of sensitivity and uncertainty analysis in an integrated manner within the framework of the BEPU approach [2,8]. Non-parametric methods have been proposed with a similar purpose, such as the Bootstrap method introduced in reference [26].…”

Section: Wilks' Methodsmentioning

confidence: 99%

“…It is possible to obtain confidence bounds of percentiles from the bootstrap sampling distribution. Different methods have been proposed in the literature to obtain confidence bounds from the bootstrap, for example, percentile method, bias-corrected accelerated method, approximated bootstrap confidence interval, bootstrap-t method or a combination of Ensemble-Based Sensitivity Analysis and the Bootstrap method [26].…”

Section: Bootstrap Methodsmentioning

confidence: 99%

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Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Sánchez-Saez

Sánchez

Sancho

et al. 2018

Reliability Engineering & System Safety

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The safety analysis of nuclear power plant is moving towards a realistic approach in which the simulations performed using best estimate computer codes must be accompanied by an uncertainty analysis, known as the Best Estimate Plus Uncertainties approach. The most popular statistical method used in these analyses is the Wilks' method, which is based on the principle of order statistics for determining a certain coverage of the Figures-of-Merit with an appropriate degree of confidence. However, there exist other statistical techniques that could provide similar or even better results. This paper explores the performance of alternative nonparametric methods as compared to the Wilks' method of obtaining such Figure-of-Merits tolerance intervals. Three methods are investigated, i.e. Hutson and Beran-Hall methods and a bootstrap method. All the techniques have been used to perform the uncertainty analysis of a Large-Break Loss of Coolant Accident. The Figure-of-Merit of interest in this application is the maximum value reached by the Peaking Clad Temperature. In order to analyze the results obtained by the different methods, four performance metrics are proposed to measure the coverage, dispersion, conservativeness, and robustness of the tolerance intervals.

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Section: Wilks' Methodsmentioning

confidence: 99%

Section: Bootstrap Methodsmentioning

confidence: 99%

Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Sánchez-Saez

Sánchez

Sancho

et al. 2018

Reliability Engineering & System Safety

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“…The substitution of the thermal-hydraulic code by metamodels is a widespread technique, which allows performing a larger number of runs with a reasonable computational cost. Some of the metamodels employed in nuclear safety analysis are, for example, finite mixture used in (Carlos et al, 2013) , (Di Maio et al, 2015), (Di Maio et al, 2016) for determining PCT. In this work, a GAM is proposed to undertake the uncertainty propagation analysis.…”

Section: Metamodel I Generalized Additive Model (Gam)mentioning

confidence: 99%

“…This technique permits to obtain the STL with a determined coverage/confidence value with reduced number of simulations (Beal, 2012), but only provides the safety variable bound. Other techniques, such as, the construction of surrogate models, or metamodelling, in which the TH code is substituted by a model in order to obtain an estimate of the FOMs of interest, or a related variable, permits extracting more information from the PDF with a reasonable computational cost (Carlos et al, 2013) (Di Maio et al, 2015) (Di Maio et al, 2016). However, as the number of input parameters increase, the construction of a metamodel requires a larger number of TH code simulations.…”

Section: Introductionmentioning

confidence: 99%

Uncertainty analysis of PKL SBLOCA G7.1 test simulation using TRACE with Wilks and GAM surrogate methods

Sánchez-Saez

Alberola

Sancho

et al. 2017

Nuclear Engineering and Design

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The Nuclear Energy Agency auspices simulation of experiments in different facilities under several programs. One on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident. The results obtained by TRACE code for PKL experiment SBLOCA G7.1 (a scaled model of Konvoi reactor) were in good agreement with the experiments. However, in the simulation process, uncertainty was not accounted. Uncertainty analysis, following the principles of Best Estimate Plus Uncertainty (BEPU) approach, must be performed to measure the effect of uncertainties on the evolution of safety variables of interest, such as the maximum of the Peak Cladding Temperature (PCTmax) in the experiment. In this paper we present a comparison between two uncertainty analysis techniques. The first technique is based on order statistics that makes use of Wilks' formula. The second technique is based on a generalized additive model (GAM) that substitutes the thermal-hydraulic code, without and with consideration of errors in adjusting the GAM model. The comparison of the uncertainty analysis results makes use of several performance metrics such as coverage, coefficient of variation and conservativeness. Based on the results of these metrics it can be concluded that the GAMPE (GAM Plus Error) provides the best performance, in particular, when using small sample size, i.e. n=59, 93. For larger sample sizes, i.e. n=124, 153, GAMPE and Wilks' results presents similar performance.

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“…Among these, OS is popular and consolidated because it provides relatively conservative results with a few computer code runs, for leveraging the usually expensive computational cost of simulation codes (Nutt and Wallis, 2004;Zio et al, 2010;Sanchez-Saez et al, 2017). In this study, we, thus, take a "Bracketing" OS approach for tackling the computational problem and calculating the safety margins (Nutt and Wallis, 2004;Di Maio et al, 2016a;Di Maio et al, 2016b;Di Maio et al, 2017).…”

Section: Introductionmentioning

confidence: 99%

A Monte Carlo-based exploration framework for identifying components vulnerable to cyber threats in nuclear power plants

Wang

Cammi

Maio

et al. 2018

Reliability Engineering & System Safety

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Bootstrapped-ensemble-based Sensitivity Analysis of a trace thermal-hydraulic model based on a limited number of PWR large break loca simulations

Abstract: The safety verification of nuclear systems can be done by analyzing the outputs of Best-Estimate Thermal-Hydraulic (BE-TH)

Cited by 14 publications

References 40 publications

Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Uncertainty analysis of PKL SBLOCA G7.1 test simulation using TRACE with Wilks and GAM surrogate methods

A Monte Carlo-based exploration framework for identifying components vulnerable to cyber threats in nuclear power plants

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