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Development of Risk Oriented Accident Analysis Methodology for Assessment of Effectiveness of Severe Accident Management Strategy in Nordic BWR
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
2019 (English)Doctoral thesis, comprehensive summary (Other academic)
Abstract [en]

Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability as a severe accident management strategy (SAM). In case of a severe accident, the debris ejected from the vessel are expected to fragment, quench and form a debris bed, which is coolable by a natural circulation of water. Success of the existing SAM strategy depends on melt release conditions from the vessel which determine (i) properties of ejected debris and, thus, ex-vessel debris bed coolability, and (ii) potential for energetic melt-coolant interactions (steam explosion). The strategy involves complex interactions between physical phenomena (deterministic) and transient accident scenarios (probabilistic).The aim of this work is further extension, implementation and application of the Risk-Oriented Accident Analysis Methodology (ROAAM) to assessment of the severe accident management strategy effectiveness. ROAAM was originally developed for rare, high-consequence hazards, where both aleatory (stochastic) and epistemic (modeling) uncertainties play a significant role in the risk assessment. The main purpose of ROAAM is to provide the input material to an underlying decision making regarding current safety design acceptance, procedures and possible design modifications.This work reports results of (i) development and implementation of probabilistic framework (ROAAM+) for streamlining sensitivity analysis, uncertainty quantification and risk analysis; (ii) analysis of in-vessel phase of accident progression and melt release conditions in Nordic BWR reactor design with MELCOR code; (iii) analysis of the effect of melt release conditions predicted by MELCOR code on the risk of ex-vessel steam explosion.In ROAAM+, “full models”, such as MELCOR code, are used to develop computationally efficient “surrogate models” to enable extensive uncertainty quantification and failure domain analysis. ROAAM+ analysis identified specific assumptions in MELCOR models, which are currently the major contributors to the uncertainty in the assessment of the SAM effectiveness.

Abstract [sv]

Den generiska ABB-reaktorn (Nordic BWR) använder inneslutningkyling, tryckavlastning och filtrering av utsläpp som strategi för hantering av svåra haverier. Vid ett svårt haveri kommer härdgrus falla ned i nedre primärutrymmet, fragmentera, och att bilda en s.k. grusbädd där resteffekten kan kylas ned med hjälp av naturlig cirkulation av vattnet i bassängen. Framgången med den befintliga strategin beror på härdsmälteförloppet och härdsmältfrigöring från reaktortanken som bestämmer förutsättningarna för: (i) egenskaper för reaktorgruset och dämed även grusbädden, och (ii) ångexplosioner som kan inträffa när härdsmältan faller ned i nedre primärutrymmet.Strategin är konceptuellt enkel, men den innebär komplexa interaktioner mellan fysiska fenomenen och processer, och är mycket känslig för olycksscenarierna. Den kan inte bedömas med hjälp av separerata probabilistiska eller deterministiska metoder på grund av osäkerhet som uppkommer från interaktioner mellan olycksscenarierna och deterministiska fenomen.Därför har så kallad Risk Oriented Accident Analysis Methodology (ROAAM) som kombinerar probabilistiska med deterministiska metoder föreslagits som riskvärdering och bedömning huruvida strategin ger ett tillräckligt skydd för omgivningen. Denna metodologi (ROAAM) utvecklades för bedömning av sällsynta högkonsekventa händelser där både aleatoriska (stokastiska) och epistemiska (modelleringsrelaterade) osäkerheter spelar en viktig roll i riskbedömningen.Huvudsyftet med ROAAMs användning är att ge indata för ett underliggande beslutsproblem och möjliggöra robust beslutsfattande gällande nuvarande säkerhetsdesign och procedurer samt möjliga konstruktionsändringar.Detta arbete är inriktat på vidareutveckling av ROAAM-metodologin, som innefattar (i) utveckling och genomförande av probabilistiska ramar för riskanalys och kvantifiering i ROAAM+; (ii) analys av svår haveriutveckling i reaktortanken, härdsmälteförloppet och förutsättningarna för härdsmältfrigöring från reaktortank som analyserats med koden MELCOR; och (iii) riskvärdering av ångexplosion i reaktorinneslutning beroende på förutsättningarna för härdsmältfrigöring från reaktortank.I ROAAM+ används "fullmodeller", såsom MELCOR-koden, för att utveckla beräkningseffektiva "surrogatmodeller" för att möjliggöra omfattande analys av osäkerhetsfaktorer och identifiera skadedomäner. ROAAM+ analys identifierade specifika antaganden i MELCOR-modeller, som för närvarande är de viktigaste bidragsgivarna till osäkerheten i bedömningen av SAM-effektiviteten.

Place, publisher, year, edition, pages
Stockholm: KTH Royal Institute of Technology, 2019. , p. 77
Series
TRITA-SCI-FOU ; 2019:08
Keywords [en]
Severe accident management, sensitivity, uncertainty, MELCOR, ROAAM
Keywords [sv]
Svår haverihantering, känslighet och osäkerhetsanalys, MELCOR, ROAAM
National Category
Engineering and Technology
Research subject
Physics
Identifiers
URN: urn:nbn:se:kth:diva-242353ISBN: 978-91-7873-103-9 (print)OAI: oai:DiVA.org:kth-242353DiVA, id: diva2:1283939
Public defence
2019-02-27, FA31, Roslagstullsbacken 21, Stockholm, 13:00 (English)
Opponent
Supervisors
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
List of papers
1. A framework for assessment of severe accident management effectiveness in Nordic BWR plants
Open this publication in new window or tab >>A framework for assessment of severe accident management effectiveness in Nordic BWR plants
Show others...
2014 (English)In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper, Published paper (Refereed)
Abstract [en]

In the case of severe accident in Nordic boiling water reactors (BWR), core melt is poured into a deep pool of water located under the reactor. The severe accident management (SAM) strategy involves complex and coupled physical phenomena of melt-coolant-structure interactions sensitive to the transient accident scenarios. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) if corium debris bed is coolable, and (ii) potential for energetic steam explosion. The goal of this work is to develop a risk-oriented accident analysis framework for quantifying conditional threats to containment integrity for a Nordic-type BWR. The focus is on the process of refining the treatment and components of the framework to achieve (i) completeness, (ii) consistency, and (iii) transparency in the review of the analysis and its results. A two-level coarse-fine iterative refinement process is proposed. First, fine-resolution but computationally expensive methods are used in order to develop computationally efficient surrogate models. Second, coupled modular framework is developed connecting initial plant damage states with respective containment failure modes. Systematic statistical analysis is carried out to identify the needs for refinement of detailed methods, surrogate models, data and structure of the framework to reduce the uncertainty, and increase confidence and transparency in the risk assessment results.

National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:kth:diva-164083 (URN)2-s2.0-84925071190 (Scopus ID)
Conference
12th International Probabilistic Safety Assessment and Management Conference, PSAM 2014; Sheraton WaikikiHonolulu; United States; 22 June 2014 - 27 June 2014
Note

QC 20150416

Available from: 2015-04-13 Created: 2015-04-13 Last updated: 2019-01-30Bibliographically approved
2. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR
Open this publication in new window or tab >>Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR
2016 (English)In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 125-141Article in journal (Refereed) Published
Abstract [en]

Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario parameters. Pattern analysis is employed in order to characterize typical behavior of core relocation transients. Clustering analysis is employed for grouping of different accident scenarios, which result in similar core relocation behavior and properties of the debris.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-200215 (URN)10.1016/j.nucengdes.2016.09.029 (DOI)000390736400011 ()2-s2.0-84993993448 (Scopus ID)
Note

QC 20170202

Available from: 2017-02-02 Created: 2017-01-23 Last updated: 2019-01-30Bibliographically approved
3. Sensitivity analysis of debris properties in lower plenum of a Nordic BWR
Open this publication in new window or tab >>Sensitivity analysis of debris properties in lower plenum of a Nordic BWR
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 374-382Article in journal (Refereed) Published
Abstract [en]

Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. The melt released from the vessel is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. Melt release conditions are recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena, subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures (such as instrumentation guide tubes IGTs, control rod guide tubes CRGTs), vessel failure and melt release. This work is focused on the evaluation of uncertainty in core degradation progression and its effect on the resultant properties of relocated debris in lower plenum of Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize the range of possible debris properties in lower plenum and its sensitivity towards different modelling parameters, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the risk assessment framework.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Severe accident, Nordic BWR, ROAAM, MELCOR
National Category
Other Chemistry Topics
Identifiers
urn:nbn:se:kth:diva-227209 (URN)10.1016/j.nucengdes.2018.03.029 (DOI)000430395700033 ()
Note

QC 20180529

Available from: 2018-05-29 Created: 2018-05-29 Last updated: 2019-01-30Bibliographically approved
4. Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
Open this publication in new window or tab >>Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
2019 (English)In: Science and Technology of Nuclear InstallationsArticle in journal (Refereed) Submitted
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-242349 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
5. Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR
Open this publication in new window or tab >>Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR
2019 (English)In: Nuclear Engineering and DesignArticle in journal (Refereed) Submitted
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242350 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
6. Implementation of Probabilistic Framework of Risk Analysis Framework for Assessment of Severe Accident Management Effectiveness in Nordic BWR
Open this publication in new window or tab >>Implementation of Probabilistic Framework of Risk Analysis Framework for Assessment of Severe Accident Management Effectiveness in Nordic BWR
(English)In: Annals of Nuclear EnergyArticle in journal (Refereed) Submitted
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242351 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
7. Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
Open this publication in new window or tab >>Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
(English)In: Annals of Nuclear EnergyArticle in journal (Refereed) Submitted
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242352 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
8. The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR
Open this publication in new window or tab >>The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR
2019 (English)Conference paper, Published paper (Refereed)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242346 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
9. Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code
Open this publication in new window or tab >>Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code
2019 (English)Conference paper, Published paper (Refereed)
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-242348 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved

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5678910118 of 17
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