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Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
2016 (engelsk)Inngår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Konferansepaper, Publicerat paper (Fagfellevurdert)
Abstract [en]

In a previous study [1], the US NRC code TRACE was employed to simulate the FIX-II tests which were carried out to investigate the loss of coolant accident (LOCA) of a boiling water reactor (BWR). Results exhibited that the TRACE simulation was sensitive to modelling parameters. In order to further qualify the TRACE code for BWR safety analysis, and to increase our confidence in the simulation results, sensitivity and uncertainty analysis is performed in this paper for the possible uncertain parameters, so as to identify the most influential ones. 12 parameters related to the simulated physical phenomena are selected by resorting to phenomena identification and ranking tables (PIRTs) in relative references. The sensitivity analysis method chosen is based on Finite Mixture Models (FMM) together with Hellinger distance and Kullback-Leibler divergence. Kolmogorov-Smirnov test is first introduced to combine FMM, and it has better performance in screening. Sensitivity analysis results of FMM method show that decay power, choked flow multipliers and break area have the most important influence on calculating peak cladding temperature (PCT). Although previous study failed to predict PCT, uncertainty analysis provides a certain range that successfully covers experiment result.

sted, utgiver, år, opplag, sider
Association for Computing Machinery (ACM), 2016.
Emneord [en]
LOCA, Sensitivity, Thermal-hydraulics simulation, TRACE, Uncertainty analysis
HSV kategori
Identifikatorer
URN: urn:nbn:se:kth:diva-234521Scopus ID: 2-s2.0-85052524461OAI: oai:DiVA.org:kth-234521DiVA, id: diva2:1246443
Konferanse
17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Qujiang Int'l Conference CenterXi'an, Shaanxi, China, 3 September 2017 through 8 September 2017
Merknad

QC 20180907

Tilgjengelig fra: 2018-09-07 Laget: 2018-09-07 Sist oppdatert: 2018-09-07bibliografisk kontrollert

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