2018
DOI: 10.1016/j.ress.2018.02.005
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Uncertainty analysis of a large break loss of coolant accident in a pressurized water reactor using non-parametric methods

Abstract: 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 othe… Show more

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Cited by 49 publications
(14 citation statements)
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“…The P.O.F is then optimized globally using a differential evolution solver [29]. Algorithm 1 summarizes step 1 to step 4 in order to compute the lowest probability of failure (4). The main cost of the algorithm arises from the high number of metamodel calls for G, evaluated on a d-dimensional grid of size ) and its corresponding sequences of canonical moments, p i = (p…”
Section: Step 4 Computation Of the Objective Functionmentioning
confidence: 99%
“…The P.O.F is then optimized globally using a differential evolution solver [29]. Algorithm 1 summarizes step 1 to step 4 in order to compute the lowest probability of failure (4). The main cost of the algorithm arises from the high number of metamodel calls for G, evaluated on a d-dimensional grid of size ) and its corresponding sequences of canonical moments, p i = (p…”
Section: Step 4 Computation Of the Objective Functionmentioning
confidence: 99%
“…Indeed, measured data possess an intrinsic level of uncertainty, due to its evolutionary nature (i.e. time-space variability) and unavoidable measurement imprecision or even errors [20]. In the case of real-time monitoring, the degree of uncertainty affecting the prediction…”
Section: Accepted Manuscriptmentioning
confidence: 99%
“…One precise case is where operators have to study loss of coolant accidents (LOCA), which result in a break in the primary loop of pressurized water nuclear reactors. This scenario can be simulated using system thermal-hydraulic computer codes which involve dozens of physical parameters including such things as condensation and heat transfer coefficients (Mazgaj et al, 2016;Sanchez-Saez et al, 2018). However, the values of many such parameters are known with only limited precision (Larget, 2019) as they are typically calculated by way of other quantities measured via small-scale physical experiments.…”
Section: Introductionmentioning
confidence: 99%