Change search
CiteExportLink to record
Permanent link

Direct link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf
Application of integrated deterministic-probabilistic safety analysis to assessment of severe accident management effectiveness in Nordic BWRs
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 Power Safety.
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
Show others and affiliations
2016 (English)In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper, Published paper (Refereed)
Abstract [en]

The goal of this work is to assess effectiveness of severe accident management strategy in Nordic type boiling water reactors (BWRs). Corium melt released into a deep pool of water below reactor vessel is expected to be fragmented to form a porous debris bed coolable by natural circulation of coolant. However, there is a risk that energetic steam explosion or non-coolable debris can threaten containment integrity. Both stochastic accident scenario (aleatory) and modeling (epistemic) uncertainties contribute to the risk assessment. Namely, the effects of melt release characteristics (jet diameter, melt composition, superheat), water pool conditions (i.e. depth and subcooling) at the time of the release, and modeling assumptions have to be quantified in a consistent manner. In order to address the uncertainty, we develop a Risk Oriented Accident Analysis framework (ROAAM+) where all stages of the accident progression are simulated using a set of models coupled through initial and boundary conditions. The analysis starts from plant damage states determined in PSA Level-1 and follows time dependent accident scenarios 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 development of simplified but computationally efficient surrogate models (SM), we employ databases of solutions obtained by detailed but computationally expensive full models (FM). The process includes iterative refining of the framework, full and surrogate models in order to achieve completeness, consistency, and transparency in the review of the analysis results. 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
Association for Computing Machinery (ACM), 2016.
Keywords [en]
BWR, ROAAM, Severe Accident
National Category
Other Civil Engineering
Identifiers
URN: urn:nbn:se:kth:diva-234525Scopus ID: 2-s2.0-85052530024OAI: oai:DiVA.org:kth-234525DiVA, id: diva2:1246361
Conference
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
Note

QC 20180907

Available from: 2018-09-07 Created: 2018-09-07 Last updated: 2018-09-07Bibliographically approved

Open Access in DiVA

No full text in DiVA

Scopus

Authority records BETA

Kudinov, PavelGrishchenko, DmitryBasso, Simone

Search in DiVA

By author/editor
Kudinov, PavelGalushin, SergeyGrishchenko, DmitryYakush, SergeyBasso, SimoneKonovalenko, AlexanderDavydov, Mikhail
By organisation
Nuclear Power Safety
Other Civil Engineering

Search outside of DiVA

GoogleGoogle Scholar

urn-nbn

Altmetric score

urn-nbn
Total: 24 hits
CiteExportLink to record
Permanent link

Direct link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf