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Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.ORCID iD: 0000-0002-0683-9136
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.ORCID iD: 0000-0001-8216-9376
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
I nstitute for Problems in Mechanics, Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow, 119526, Russian Federation.
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, p. 2519-2535Conference paper, Published paper (Refereed)
Abstract [en]

In this work we present results of development and application of Risk Oriented Accident Analysis framework (ROAAM+) to assessment of effectiveness of ex-vessel severe accident management strategy. In case of a core melt accident in Nordic type boiling water reactor (BWR) corium is released into a deep pool of water below reactor vessel to form a porous bed of debris. Energetic steam explosion or formation of non-coolable debris can threaten containment integrity. Both stochastic (aleatory) accident scenario and modeling (epistemic) uncertainties contribute to uncertainty. ROAAM+ framework is developed to simulate the whole accident progression The analysis starts from plant damage states determined in PSA Level-1 and continues with analysis of core degradation, vessel failure, melt release, steam explosion and debris bed formation and coolability. In order to achieve computational efficiency sufficient for extensive sensitivity, uncertainty, and risk analysis the surrogate modeling approach is used. In the paper we present results of the analysis aimed at quantification of uncertainty in the conditional containment failure probability. Specifically, we carry out sensitivity analysis using standalone and coupled models in order to identify the most influential scenario and modeling parameters for each sub-model. We assess the impact of the parameters on the prediction of the “load”, “capacity” and also failure probability. Then we quantify the effect of the most influential parameters on the failure probability. The results are presented using the failure domain approach and second order probability analysis, considering the uncertainty in distributions of the input parameters.

Place, publisher, year, edition, pages
2019. p. 2519-2535
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:kth:diva-268324Scopus ID: 2-s2.0-85073714269OAI: oai:DiVA.org:kth-268324DiVA, id: diva2:1413394
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019; Marriott Portland Downtown WaterfrontPortland; United States; 18 August 2019 through 23 August 2019
Note

QC 20200310

Available from: 2020-03-10 Created: 2020-03-10 Last updated: 2020-03-10Bibliographically approved

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Kudinov, PavelGalushin, SergeyGrishchenko, Dmitry

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