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  • 1.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Development of Risk Oriented Accident Analysis Methodology for Assessment of Effectiveness of Severe Accident Management Strategy in Nordic BWR2019Doctoral 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.

  • 2.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Implementation of Probabilistic Framework of Risk Analysis Framework for Assessment of Severe Accident Management Effectiveness in Nordic BWRIn: Annals of Nuclear EnergyArticle in journal (Refereed)
  • 3.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Risk analysis framework for severe accident mitigation strategy in nordic BWR: An approach to communication and decision making2017In: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, American Nuclear Society , 2017, p. 587-594Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWRs) employ ex-vessel debris cooling in a deep water pool. The success of the strategy requires (i) formation of a coolable porous debris bed; (ii) no energetic steam explosion that can threaten containment integrity. Both scenario (aleatory) and modeling (epistemic) uncertainties are important in the assessment of the failure risks. A consistent approach is necessary for the decision making on whether the strategy is sufficiently effective, or a modification of the SAM is necessary. Risk Oriented Accident Analysis Methodology (ROAAM+) is a tool for assessment of failure probability to enable robust decision making, insensitive to remaining uncertainty. Conditional containment failure probability is considered in this work as an indicator of severe accident management effectiveness for Nordic BWR. The ultimate goal of ROAAM+ application for Nordic BWR is to provide a scrutable background in order to achieve convergence of experts' opinions in decision making. The question is: if containment failure can be demonstrated as physically unreasonable, given severe accident management strategy and state-of-the-art knowledge? If inherent safety margins are large, then the answer to the question is positive and can be demonstrated through risk assessment with consistent conservative treatment of uncertainties and by improving, when necessary, knowledge and data. Otherwise, the risk management should be applied in order to increase margins and achieve the safety goal through modifications of the SAM (e.g. safety design, SAMGs, etc.). The challenge for a decision maker is to distinguish when collecting more knowledge and reduction of uncertainty in risk assessment or application of risk management with SAM modifications would be the most effective and efficient approach. In this work we demonstrate a conceptual approach for communication of ROAAM+ framework analysis results and provide an example of a decision support model. The results of the risk analysis are used in order to provide necessary insights on the conditions when suggested changes in the safety design are justified.

  • 4.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code2019Conference paper (Refereed)
  • 5.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    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 BWR2019Conference paper (Refereed)
  • 6.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 125-141Article in journal (Refereed)
    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.

  • 7.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR2019In: Nuclear Engineering and Design, Vol. 350, p. 243-258Article in journal (Refereed)
    Abstract [en]

    Nordic Boiling Water Reactors (BWRs) rely on the flooding of the lower drywell (LDW) as a severe accident management (SAM) strategy. The termination of a SA is achieved by fragmenting and quenching of the melt released from the vessel. Success of SAM strategy depends on melt release and water pool conditions. The characteristics of the melt release are the major source of uncertainty in quantification of the risk of SAM failure. Vessel failure and melt release modes are subject to aleatory and epistemic uncertainties at the in-vessel accident progression stage. In this work we focus on predicting the properties of debris relocated to the lower plenum using MELCOR code. We address the effect of epistemic uncertainty in modeling parameters and models in the MELCOR code in different severe accident scenarios on main characteristics of the in-vessel accident progression in Nordic BWRs. Sensitivity analysis is performed to rank the importance of MELCOR modelling parameters and the effect of different MELCOR models is addressed by using different versions of the code. The results provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on code predictions.

  • 8.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code2019In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 5310808Article in journal (Refereed)
    Abstract [en]

    Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

  • 9.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code2019In: Science and Technology of Nuclear InstallationsArticle in journal (Refereed)
  • 10.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, melt is 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 natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties and thus coolability of the debris bed, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. Melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs using ROAAM+ Framework. The melt release conditions, including in-vessel\ex-vessel pressure, lower drywell pool depth and temperature, are affected by aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we use MELCOR code to perform the analysis of the effects of Severe accident scenarios and modelling options in MELCOR on the properties of debris relocated to the lower head, the time and the mode of vessel lower head failure. We identify the most influential uncertain factors and discuss the needs for improvements in the modeling approaches. 

  • 11.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Comparison of melcor code versions predictions of the properties of relocated debris in lower plenum of nordic BWR2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt 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 vessel failure and melt release mode from the vessel, which determine conditions for (i) the formation of debris bed and its coolability, and (ii) steam explosion. Non-coolable debris and strong explosions present credible threats to containment integrity. A risk oriented accident analysis framework (ROAAM+) is under development for assessment of the effectiveness of the severe accident management strategy. 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, vessel failure and melt release. In this work we perform comparison of predictions of different MELCOR code versions used for the analysis of the effect severe accident scenario and uncertainties on the processes of core degradation and relocation, and resulting properties of relocated debris in Nordic BWR lower plenum. Properties of relocated debris are obtained as functions of the accident scenario parameters, such as timing of activation of different safety systems. We perform the analysis of the codes predictions and discuss possible reasons for the discrepancies in observations. The main goal of this work is to provide insights regarding the effect of code uncertainty, sensitivity coefficients and user effect on the code predictions, which is of importance for the analysis of in-vessel debris coolability and vessel failure mode in the ROAAM+ framework.

  • 12.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effect of severe accident scenario and modeling options in melcor on the properties of relocated debris in nordic BWR lower plenum2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt 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 vessel failure and melt release mode from the vessel, which determine conditions for (i) the formation of debris bed and its coolability, and (ii) steam explosion. Non-coolable debris and strong explosions present credible threats to containment integrity. A risk oriented accident analysis framework (ROAAM+) is under development for assessment of the effectiveness of the severe accident management strategy. 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, 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 ROAAM+ framework.

  • 13.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Scenario Grouping and Classification Methodology for Postprocessing of Data Generated by Integrated Deterministic-Probabilistic Safety Analysis2015In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 278638Article in journal (Refereed)
    Abstract [en]

    Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) combines deterministic model of a nuclear power plant with a method for exploration of the uncertainty space. Huge amount of data is generated in the process of such exploration. It is very difficult to "manually" process and extract from such data information that can be used by a decision maker for risk-informed characterization, understanding, and eventually decision making on improvement of the system safety and performance. Such understanding requires an approach for interpretation, grouping of similar scenario evolutions, and classification of the principal characteristics of the events that contribute to the risk. In this work, we develop an approach for classification and characterization of failure domains. The method is based on scenario grouping, clustering, and application of decision trees for characterization of the influence of timing and order of events. We demonstrate how the proposed approach is used to classify scenarios that are amenable to treatment with Boolean logic in classical Probabilistic Safety Assessment (PSA) from those where timing and order of events determine process evolution and eventually violation of safety criteria. The efficiency of the approach has been verified with application to the SARNET benchmark exercise on the effectiveness of hydrogen management in the containment.

  • 14.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Sensitivity analysis of debris properties in lower plenum of a Nordic BWR2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 374-382Article in journal (Refereed)
    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.

  • 15.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Royal Institute of Technology.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Sensitivity Analysis of the Vessel Lower Head Failure in Nordic BWR using MELCOR Code2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, 2018Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, corium is 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 natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties of the debris bed and thus if the bed is coolable or not, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. It is currently recognized that the time and the mode of vessel failure, melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in ROAAM+ Framework. The properties of relocated debris, time and the mode of vessel failure and melt release conditions, including in-vessel/ex-vessel pressure, lower drywell pool depth and temperature, are subject to aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we perform sensitivity analysis for a set of representative cases, to evaluate the effect of MELCOR modelling parameters on the process of core degradation and relocation, and vessel failure mode. Major contributors to the uncertainty in the timing of the vessel failure and amount of melt available for release at the time of failure are identified and discussed in detail. 

  • 16.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 135, article id 106976Article in journal (Refereed)
    Abstract [en]

    Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

  • 17.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effect of severe accident scenarios on in-vessel debris relocation in Nordic BWRs2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 1957-1970Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactor (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of this strategy is contingent upon melt release mode from the vessel, which determine conditions for (i) 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 in-vessel accident scenarios and phenomena, subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine the conditions for corium interactions with vessel structures, vessel failure and melt release. This work is focused on the evaluation of the effect of a severe accident scenario on the process of core degradation progression and resultant properties of relocated debris in the lower plenum of a Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize possible debris properties in the lower plenum and its sensitivity to severe accident scenario parameters, such as performance of safety systems and possible operator actions, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the risk assessment framework.

  • 18.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. employer.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Risk Analysis Framework for Decision Support for Severe Accident Mitigation Strategy in Nordic BWR2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, 2018Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWRs) employ ex-vessel debris cooling in a deep water pool. The success of the strategy requires formation of a coolable porous debris bed; no energetic steam explosion that can threaten containment integrity. Both scenario (aleatory) and modeling (epistemic) uncertainties are important in the assessment of the failure risks. A consistent approach is necessary for the decision making on whether the strategy is sufficiently effective, or modification of the SAM is necessary. Risk Oriented Accident Analysis Methodology (ROAAM+) is a tool for assessment of failure probability to enable robust decision making, insensitive to remaining uncertainty. The challenge for a decision maker is to distinguish the cases when collecting more knowledge and reduction of uncertainty in risk assessment, or modification of risk management strategy would be the most adequate approach given the safety goals and criteria. When either decision is made, ROAAM+ can provide data for selection of the most efficient implementation of the decision by selecting research priorities or modifying design elements that contribute most to the risk. In this work we discuss different approaches for communication of ROAAM+ framework analysis results and decision support. We focus on connection and integration of ROAAM+ results into risk-informed decision making models used in nuclear industry. The results of risk analysis are used in order to provide necessary insights on conditions when suggested changes in the safety design can be justified, taking into account different aspects of risk.

  • 19.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Ranlöf, L.
    Bäckström, O.
    Adolfsson, Y.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Marklund, A. R.
    Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper (Refereed)
    Abstract [en]

    A comprehensive and robust assessment of severe accident management effectiveness in preventing unacceptable releases is a challenge for a today’s real life PSA. This is mainly due to the fact that major uncertainty is determined by the physical phenomena and timing of the events. The static PSA is built on choosing scenario parameters to describe the accident progression sequence and typically uses a limited number of simulations in the underlying deterministic analysis. Risk Oriented Accident Analysis Methodology framework (ROAAM+) is being developed in order to enable consistent and comprehensive treatment of both epistemic and aleatory uncertainties. The framework is based on a set of deterministic models that describe different stages of the accident progression. The results are presented in terms of distributions of conditional containment failure probabilities for given combinations of the scenario parameters. This information is used for enhanced modeling in the PSA-L2. Specifically, it includes improved definitions of the sequences determined by the physical phenomena rather than stochastic failures of the equipment, improved knowledge of timing in sequences and estimation of probabilities determined by the uncertainties in the phenomena. In this work we present an example of application of the dynamic approach in a large scale PSA model and show that the integration of the ROAAM+ results and the PSA model can potentially lead to a considerable change in PSA Level 2 analysis results. 

  • 20.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Krčál, Pavel
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Ranlöf, Lisa
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Bäckström, Ola
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Adolfsson, Yvonne
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An approach to joint application of integrated deterministic-probabilistic safety analysis and PSA Level 2 to severe accident issues in Nordic BWRs2016In: PSAM 2016 - 13th International Conference on Probabilistic Safety Assessment and Management2017, International Association for Probabilistic Safety Assessment and Management (IAPSAM), 2016Conference paper (Refereed)
    Abstract [en]

    In this paper we outline a conceptual approach for combined use of Probabilistic Safety Assessment (PSA) and Integrated Deterministic-Probabilistic Safety Assessment (IDPSA), considering Nordic Boiling Water Reactor (BWR) severe accident issues (specifically ex-vessel steam explosion and debris bed coolability) for illustration.

    We describe a conceptual approach based on post processing of the results generated by IDPSA to update the “static” Boolean structures in the standard PSA representation. The challenge in the evaluation is to retain the failure combinations from the PSA to allow for component importance evaluation, to be able to perform the calculations in a reasonable time frame and to use all relevant information from the IDPSA results.

    We discuss the approaches for determination of the event space (for IDPSA analysis) which is consistent with PSA damage states from PSA-L1 and L2. We also discuss application of post processing approach for analysis of huge amount of data generated in the process of uncertainty space exploration, which is difficult to use directly in decision making process including incorporation of such data into PSA framework, to update structures of “static” Boolean structures in standard PSA. Using data post processing approach we can significantly reduce amount of information which represents results from IDPSA analysis.

  • 21.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Approach to Grouping and Classification of Scenarios in Integrated Deterministic-Probabilistic Safety Analysis2014In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper (Other academic)
    Abstract [en]

    Integrated Deterministic Probabilistic Safety Assessment (IDPSA) methodologies aim to achieve completeness and consistency of the analysis. However, for the purpose of risk informed decision making it is often insufficient to merely calculate a quantitative value for the risk and its associated uncertainties. IDPSA combines deterministic model of a nuclear power plant with a method for exploration of the uncertainty space. Huge amount of data is generated usually in the process of such exploration. It is very difficult to "manually" process and extract from such data information that can be used by a decision maker for risk-informed characterization and eventually improvement of the system safety and performance. Such understanding requires an approach to the interpretation, grouping of similar scenario evolutions, and classification of the principal characteristics of the events that contribute to the risk. In this work we develop an approach for classification and characterization of failure domains (domains of uncertain parameters where critical system parameters exceed safety thresholds). The method is based on scenario grouping and clustering with application of decision trees for characterization of the influence of timing and order of the events

  • 22.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of Core Relocation Surrogate Model for Prediction of Debris Properties in Lower Plenum of a Nordic BWR2016In: NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Gyeongju, Korea, October 9-13, 2016. Paper N11P1234, NUTHOS-11 , 2016, article id N11P1234Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt 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. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. A consistent treatment of these uncertainties requires Integrated Deterministic Probabilistic Safety Analysis (IDPSA). We employ the concepts and approaches described in Risk Oriented Accident Analysis Methodology (ROAAM) for development of a probabilistic framework (ROAAM+) that is based on extensive uncertainty and sensitivity analysis in risk quantification. Direct application of such fine-resolution models for extensive sensitivity and uncertainty analysis is often unaffordable. We use “surrogate models” (SMs) that provide computationally efficient approximations for the FMs. In this work we demonstrate an approach to the development of Core relocation SM based on the MELCOR code as the full model (FM). We discuss the development of the database of the FM solutions, data mining and post-processing of the results for SM development. Extensive sensitivity and uncertainty analysis is carried out using the FM and implications of the analysis are discussed in detail. We demonstrate how the connection between different stages of severe accident progression is made in ROAAM+ framework for Nordic BWRs.

  • 23.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 9, p. 7222-7235Conference paper (Refereed)
    Abstract [en]

    Severe accident mitigation strategies in Nordic boiling water reactors (BWRs) employ core melt cooling in a deep pool of water under the reactor pressure vessel. Corium melt released from the vessel is expected to fragment, solidify and form a porous debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. Significant aleatory and epistemic uncertainties exist in accident scenarios, melt release conditions, and modeling of steam explosion phenomena. Assessment of the risk of ex-vessel steam explosion requires application of the Integrated Deterministic Probabilistic Safety Analysis (IDPSA). IDPSA is a computationally demanding task which makes unfeasible direct application of Fuel-Coolant Interaction codes. The goal of the current work is to develop a Surrogate Model (SM) of the Texas-V code and demonstrate its application to the analysis of explosion impact in the Nordic BWR. The SM should be computationally affordable for IDPSA analysis. We focus on prediction of the steam explosion loads in a reference Nordic BWR design assuming a scenario of coherent corium jet release into a deep water pool. We start with the review of the Texas-V sub-models in order to identify a list of parameters to be considered in implementation of the SM. We demonstrate that Texas-V exhibits chaotic response in terms of the explosion impulse as a function of the triggering time and introduce a statistical representation of the explosion impulse for given melt release conditions and arbitrary triggering time. We demonstrate that characteristics of the distribution are well-posed. We then separate out the essential portion of modelling uncertainty by identification of the most influential uncertain parameters using sensitivity analysis. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in FCI modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a Nordic BWR. A database of Texas-V solutions is generated and used for the development of the SM. Performance, predictive capability and application of the SM to risk analysis are discussed in detail.

  • 24.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR2017In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Conference paper (Refereed)
    Abstract [en]

    Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel in order to fragment and quench core melt and provide long term cooling of the debris. One of the risk factors associated with this accident management strategy is early failure of the containment due to steam explosion. Assessment of the risk is subject to significant epistemic and aleatory uncertainties in (i) modelling of steam explosion and (ii) scenarios of melt release from the vessel and water pool conditions. High computational efficiency of the models is required for such assessment. A surrogate model (SM) approach has been previously developed using artificial neural network and the database of Texas-V code solutions for steam explosion loads in the Nordic type BWRs. In this paper we extend our surrogate model to allow analysis of steam explosion in relatively shallow water pools (>2 m), address effects of melt emissivity and resolve more accurately variation of pressure in the drywell. We provide detailed comparison of metallic vs oxidic melt release scenarios, incorporate uncertainty of the SM into modelling and analyze the sensitivity of our results to SM uncertainty. We estimate risks of containment failure with non-reinforced and reinforced hatch door and demonstrate the effect of the surrogate model uncertainty on the results. We analyze the results and develop a simplified approach for decision making considering predicted failure probabilities, expected costs and scenario frequencies. 

  • 25.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 63-75Article in journal (Refereed)
    Abstract [en]

    Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel to fragment and quench core melt and provide long term cooling of the debris. One of the risks associated with this strategy is early containment failure due to ex-vessel steam explosion. Assessment of the risk of steam explosion is subject to significant (i) epistemic uncertainties in modelling and (ii) aleatory uncertainties in scenarios of melt release. For quantification of the uncertainties and the risk a full model (FM) based on TEXAS-V code and a computationally efficient surrogate model (SM) have been previously developed. FM is used to provide a database of solutions that is used for development of a SM, while SM is used in extensive sensitivity and uncertainty analysis. In this work, we compare the risk of containment failure with non-reinforced and reinforced hatch door for metallic and oxidic melt release scenarios. We quantify the error of SM in the approximation of the FM and assess the effect of the approximation uncertainty on risk assessment. We analyze the results and suggest a simplified approach for decision making considering predicted failure probabilities, expected costs, and scenario frequencies.

  • 26.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Risk of containment failure due to ex-vessel steam explosion for Nordic BWRs2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 4032-4038Conference paper (Refereed)
    Abstract [en]

    In case of a severe accident in a Light Water Reactor (LWR) degraded core relocates into the lower head of the reactor pressure vessel. Under thermal and mechanical loads from the core debris the vessel can fail releasing hot debris into the containment. In some designs of LWRs the severe accident mitigation strategy aims to prevent early containment failure by providing a pool of water below the reactor vessel. The melt is expected to form a coolable debris configuration preventing or delaying release of radioactive materials to the environment. One of the risk factors associated with melt-water interaction is containment failure due to ex-vessel steam explosion. Energetics of the steam explosion is contingent upon characteristics of melt release, pool and containment geometry. A general purpose full and surrogate models for estimation of the steam explosion loads in various conditions prototypic to boiling and pressurized water reactors have been proposed. In this paper, we rely on our recent results in model validation to develop a new surrogate model for the estimation of the steam explosion loads in LWRs using less conservative assumptions. We sample model output using Risk Oriented Accident Analysis Methodology code (ROAAM+) and provide estimates for the risk of containment failure for Nordic BWR given different accident scenarios. We plot Failure Domain maps and discuss implication of the steam explosion for different designs (fragility levels) and severe accident management strategies (pool depths). Importantly, we analyze the effect of the reduced model conservatism on the results of the risk analysis and discuss its implications to the decision making.

  • 27.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yakush, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Davydov, Mikhail
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Application of integrated deterministic-probabilistic safety analysis to assessment of severe accident management effectiveness in Nordic BWRs2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference 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.

  • 28.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yakush, Sergey E.
    I nstitute for Problems in Mechanics, Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow, 119526, Russian Federation.
    Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, p. 2519-2535Conference 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.

  • 29.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Yakush, Sergey
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Phung, Viet-Anh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Nam
    A framework for assessment of severe accident management effectiveness in Nordic BWR plants2014In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference 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.

  • 30. Phung, Viet-Anh
    et al.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Raub, Sebastian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Goronovski, Andrei
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Koop, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Characteristics of debris in the lower head of a BWR in different severe accident scenarios2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, p. 359-370Article in journal (Refereed)
    Abstract [en]

    Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel-coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input can result in large changes in the final mass of debris. Typical ranges of the debris properties in different scenarios are discussed in detail.

  • 31.
    Phung, Viet-anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Raub, Sebastian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Goronovski, Andrei
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Prediction of Corium Debris Characteristics in Lower Plenum of a Nordic BWR in Different Accident Scenarios Using MELCOR Code2015In: 2015 International Congress on Advances in Nuclear Power Plants, Nice, France: ICAPP , 2015, , p. 11Conference paper (Refereed)
    Abstract [en]

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed configurations in the lower plenum.

  • 32.
    Phung, Viet-Anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, p. 461-476Article in journal (Refereed)
    Abstract [en]

    Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Characteristics of debris bed in the reactor vessel lower plenum affect corium interactions with the vessel structures, vessel failure and melt ejection mode. Melt release from the vessel determines conditions for ex-vessel accident progression and potential threats to containment integrity, i.e. steam explosion, debris bed formation and coolability. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles. Another problem is that code predictions can be quite sensitive to small variations of the input parameters and numerical discretization, due to the highly non-linear feedbacks and discrete thresholds employed in the code models.

     

    The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. Station black out scenario with a delayed power recovery is considered. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.

  • 33.
    Yu, Peng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR2018Conference paper (Refereed)
    Abstract [en]

    We present a coupled thermo-mechanical creep analysis for a Nordic BWR lower head with a non-homogeneous debris bed configuration generated with MELCOR code. A one-way coupling approach was adopted which uses the Phase-Change Effective Convectivity Model implemented in Fluent to simulate the convective heat transfer in the melt pool and the ANSYS Mechanical to simulate the vessel wall deformation induced by the thermal and mechanical load from the debris. An initial non-homogeneity of debris bed was estimated using MELCOR core relocation simulation results specifying the mass of each component (UO2/Zr/ZrO2/SS/SSOX) and temperature in each MELCOR cell of the lower head. A mapping scheme was designed to transfer this non-homogeneities debris bed to Fluent through User Defined Functions. All components were locally treated in Fluent as one ideal phase by averaging the weights of element-specific mass fractions inside each cell. Material properties (density, heat capacity, etc.) and volumetric heat in the debris were both spatial- and temperature-dependent. Meanwhile, additional simulations using homogeneous debris bed configuration but with the same amount of mass compositions were run for comparison. Results including temperature escalation, vessel failure timing and location were analyzed and compared.

1 - 33 of 33
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