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Sökning: WFRF:(Kudinov Pavel 1972 )

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1.
  • Basso, Simone, et al. (författare)
  • Validation of DECOSIM code against experiments on particle spreading by two-phase flows in water pool
  • 2016
  • Ingår i: Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, NUTHOS-11.
  • Konferensbidrag (refereegranskat)abstract
    • Validation simulations by DECOSIM code are performed against recent PDS-P experiments on particle spreading in a planar vertical water pool with bottom air injection. The model implemented in the code considers two-fluid formulation (water, air), turbulence effects in liquid phase are taken into account by k-epsilon model with additional generation terms accounting for two-phase effects. Particles are described by Lagrangian model, with turbulent dispersion modeled by random-walk model. Simulations are performed in conditions corresponding to experimental setup, the test section was a plane rectangular tank of variable length (0.9 and 1.5 m) and pool depth (0.5, 0.7, and 0.9 m), the superficial gas injection velocity ranged between 0.12 and 0.69 m/s. Sedimentation of spherical stainless steel (1.5 and 3 mm) and glass (3 mm) particles was calculated and compared with experiments with respect to the mean spreading distance and lateral distributions of mass fraction of particles. Reasonable agreement between the results obtained and experimental measurements is achieved for all pool geometries, gas injection rates, and particle types, confirming adequacy of the modeling approach and suitability of DECOSIM code for severe accident analysis related to debris bed formation. Possible ways to further reduction of uncertainty in model validation are discussed.
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2.
  • Estévez-Albuja, S., et al. (författare)
  • Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 136
  • Tidskriftsartikel (refereegranskat)abstract
    • Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.
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3.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Effective momentum induced by steam condensation in the oscillatory bubble regime
  • 2019
  • Ingår i: International Journal of Multiphase Flow. - : Elsevier BV. - 0301-9322 .- 1879-3533. ; 350, s. 259-274
  • Tidskriftsartikel (refereegranskat)abstract
    • The spargers used in Boiling Water Reactors (BWR) discharge steam from the primary coolant system into a pool of water. Direct steam condensation in subcooled water creates sources of heat and momentum determined by the condensation regimes, called “effective sources” in this work. Competition between the effective sources can result in thermally stratification or mixing of the pool. Thermal stratification is a safety concern in BWRs since it reduces the steam condensation and pressure suppression capacity of the pool. In this work, we present semi-empirical correlations to predict the effective momentum induced by steam condensation in the oscillatory bubble regime, relevant for the operation of spargers in BWRs. A Separate Effect Facility (SEF) was designed and built at LUT, Finland, in order to provide the necessary data. An empirical correlation for the effective momentum as a function of the Jakob number is proposed. The Kelvin Impulse theory was also applied to estimate the effective momentum based on information about the bubble dynamics. To do this, new correlations for the bubble collapse frequencies, maximum bubble radius, velocities, pressure gradient and heat transfer coefficient are proposed and compared to available data from the literature. The effective momentum induced by sonic steam jets appears to be constant in a wide range of studied Jakob number. However, further experimental data is necessary at larger Jakob numbers and steam mass fluxes.
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4.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Pool stratification and mixing during a steam injection through spargers: analysis of the PPOOLEX and PANDA experiments
  • 2018
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 337, s. 300-316
  • Tidskriftsartikel (refereegranskat)abstract
    • Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors to condense steam in large water pools. A steam injection induces heat, momentum and mass sources that depend on the steam injection conditions and can result in thermal stratification or mixing of the pool. Thermal stratification reduces the steam condensation capacity of the pool, increases the pool surface temperature and thus the containment pressure. Development of models with predictive capabilities requires the understanding of basic phenomena that govern the behavior of the complex multi-scale system. The goals of this work are (i) to analyze and interpret the experiments on steam injection into a pool through spargers performed in the large-scale facilities of PPOOLEX and PANDA, and (ii) to discuss possible modelling approaches for the observed phenomena. A scaling approach was developed to address the most important physical phenomena and regimes relevant to prototypic plant conditions. The focus of the tests was on the low steam mass flux and oscillatory bubble condensation regimes, which are expected during a long-term steam injection transient, e.g. in the case of a Station Black Out (SBO). Exploratory tests were also done for chugging and stable jet conditions. The results showed a similar behavior in PPOOLEX and PANDA in terms of jet induced by steam condensation, pool stratification, and development of hot layer and erosion of the cold one. A correlation using the Richardson number is proposed to model the erosion rate of the cold layer as a function of the pool dimensions and steam injection conditions.
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5.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Pool stratification and mixing induced by steam injection through spargers : CFD modelling of the PPOOLEX and PANDA experiments
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 347, s. 67-85
  • Tidskriftsartikel (refereegranskat)abstract
    • Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Generation III/III+ Pressurized Water Reactors (PWR) to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models are implemented in ANSYS Fluent 17.0 Computational Fluid Dynamics (CFD) code and calibrated against the spargers experiments performed in the PPOOLEX and PANDA facilities, analysed by the authors in Gallego-Marcos et al. (2018b). CFD modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the effective momentum showed an inverse proportionality to the sub-cooling. Differences between the effective momentum calibrated for PPOOLEX and PANDA are discussed. Analysis of the calculated flow above the cold stratified layer showed that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.
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6.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression Pool
  • Annan publikation (övrigt vetenskapligt/konstnärligt)abstract
    • The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.
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7.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Thermal stratification and mixing in a Nordic BWR pressure suppression pool
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 132, s. 442-450
  • Tidskriftsartikel (refereegranskat)abstract
    • The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.
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8.
  • Galushin, Sergey, et al. (författare)
  • Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 350, s. 243-258
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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9.
  • Galushin, Sergey, et al. (författare)
  • Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
  • 2019
  • Ingår i: Science and Technology of Nuclear Installations. - : Hindawi Publishing Corporation. - 1687-6075 .- 1687-6083.
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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10.
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11.
  • Galushin, Sergey, et al. (författare)
  • Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code
  • 2018
  • Ingår i: PSAM 2018 - Probabilistic Safety Assessment and Management. - : International Association for Probablistic Safety Assessment and Management (IAPSAM).
  • Konferensbidrag (refereegranskat)abstract
    • 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. 
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12.
  • Galushin, Sergey, et al. (författare)
  • Analysis of the effect of vessel failure and melt release on risk of containment failure due to ex-vessel steam explosion in nordic boiling water reactor using roaam1 framework
  • 2020
  • Ingår i: Journal of Nuclear Engineering and Radiation Science. - : ASME International. - 2332-8983 .- 2332-8975. ; 6:4
  • Tidskriftsartikel (refereegranskat)abstract
    • Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The risk-oriented accident analysis methodology (ROAAM ) is used for quantification of the risk of containment failure in Nordic BWR where melt ejection mode surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.
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13.
  • Galushin, Sergey, et al. (författare)
  • Comparison of melcor code versions predictions of the properties of relocated debris in lower plenum of nordic BWR
  • 2016
  • Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
  • Konferensbidrag (refereegranskat)abstract
    • 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.
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14.
  • Galushin, Sergey, et al. (författare)
  • Comparison of vessel failure mode and melt release conditions in unmitigated and mitigated station blackout scenarios in nordic BWR using melcor code
  • 2019
  • Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE. - : American Society of Mechanical Engineers (ASME).
  • Konferensbidrag (refereegranskat)abstract
    • Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Effectiveness of this strategy depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in ROAAM+ Framework. In this work we perform the analysis of the effect of water recovery in station blackout accident in Nordic BWR on the timing and mode of vessel failure and melt release conditions using MELCOR code. The analysis is performed using Morris method in order to evaluate the sensitivity of the vessel failure mode and melt release conditions and associated uncertainty due to user-defined modelling parameters and modelling options in MELCOR code. 
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15.
  • Galushin, Sergey (författare)
  • Development of Risk Oriented Accident Analysis Methodology for Assessment of Effectiveness of Severe Accident Management Strategy in Nordic BWR
  • 2019
  • Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
    • 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.
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16.
  • Galushin, Sergey, et al. (författare)
  • Effect of severe accident scenario and modeling options in melcor on the properties of relocated debris in nordic BWR lower plenum
  • 2016
  • Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
  • Konferensbidrag (refereegranskat)abstract
    • 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.
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17.
  • Galushin, Sergey, et al. (författare)
  • Implementation of framework for assessment of severe accident management effectiveness in Nordic BWR
  • 2020
  • Ingår i: Reliability Engineering & System Safety. - : Elsevier BV. - 0951-8320 .- 1879-0836. ; 203
  • Tidskriftsartikel (refereegranskat)abstract
    • Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by (i) formation of non-coolable debris bed, or (ii) energetic steam explosion. In order to assess the effectiveness of SAM the Risk Oriented Accident Analysis Methodology framework (ROAAM +) has been developed. The framework is further extension of the approach originally developed and applied by Prof. Theofanous and co-workers. This paper presents the implementation of ROAAM + probabilistic framework for uncertainty quantification and risk analysis. We further apply ROAAM + to the analysis of steam explosion risk in Nordic BWR assuming different state-of knowledge situations and different containment fragilities. We employ an iterative processes of knowledge refinement using risk analysis as a guiding tool in identification of the major sources of uncertainty. We estimate failure domains and discuss ROAAM + results in terms of possibility vs necessity of containment failure.
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18.
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19.
  • Galushin, Sergey, et al. (författare)
  • Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR
  • 2018
  • Ingår i: PSAM 2018 - Probabilistic Safety Assessment and Management. - : International Association for Probablistic Safety Assessment and Management (IAPSAM).
  • Konferensbidrag (refereegranskat)abstract
    • 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. 
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20.
  • Galushin, Sergey, et al. (författare)
  • Quantification of the uncertainty due to state-of-knowledge using ROAAM+ framework for Nordic BWRs
  • 2019
  • Ingår i: PSA 2019 - International Topical Meeting on Probabilistic Safety Assessment and Analysis. - : American Nuclear Society. ; , s. 834-840
  • Konferensbidrag (refereegranskat)abstract
    • Risk Oriented Accident Analysis Methodology (ROAAM+) framework for Nordic BWR is a further development of ROAAM ideas, where development and application of the framework are based on iterative processes of refinement of knowledge, and where risk analysis is used as a guiding tool in the identification of the major sources of uncertainty. ROAAM+ framework employs extended treatment of safety goals, where both “possibility” and “necessity” of containment failure are considered in the analysis. One of the most important features of the ROAAM+ treatment is risk quantification in different “state-of-knowledge” situations, e.g. where complete, partial or no probabilistic knowledge available. To assess the importance of the missing information, i.e. when no probabilistic knowledge is available, distributions of epistemic modelling parameters are considered as uncertain and sampling in the space of possible probability distributions of these parameters is performed. The main goal of this work is to demonstrate an approach for risk quantification in different state-of-knowledge situations and evaluate the effect of the selection of the distribution families and parameters characterizing distributions on risk analysis results.
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21.
  • Galushin, Sergey, et al. (författare)
  • Risk analysis framework for severe accident mitigation strategy in nordic BWR : An approach to communication and decision making
  • 2017
  • Ingår i: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017. - : American Nuclear Society. - 9781510851801 ; , s. 587-594
  • Konferensbidrag (refereegranskat)abstract
    • 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.
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22.
  • Galushin, Sergey, et al. (författare)
  • Sensitivity analysis of debris properties in lower plenum of a Nordic BWR
  • 2018
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 332, s. 374-382
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
23.
  • Galushin, Sergey, et al. (författare)
  • Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 135
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
24.
  • Galushin, Sergey, et al. (författare)
  • Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code
  • 2019
  • Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE, Volume 2019-May. - : ASME Press.
  • Konferensbidrag (refereegranskat)abstract
    • Effectiveness of severe accident management strategy in Nordic BWR reactors depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in the risk of containment failure in Nordic BWRs. The Risk Oriented Accident Analysis Methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR, which relies on extensive use of computationally efficient surrogate models (SMs) for sensitivity and uncertainty analyses in risk quantification. The surrogate models provide computationally efficient approximations for the most important parameters of the computationally expensive full models. In the ROAAM+ framework the melt ejection surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine respective loads on the containment. This paper demonstrates an approach to the development and application of the melt ejection surrogate model based on the MELCOR code results. The post-processing of the MELCOR code results was performed in order to establish a connection between different stages of severe accident progression in ROAAM+ framework for Nordic BWRs.
  •  
25.
  • Galushin, Sergey, et al. (författare)
  • The effect of severe accident scenarios on in-vessel debris relocation in Nordic BWRs
  • 2019
  • Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 1957-1970
  • Konferensbidrag (refereegranskat)abstract
    • 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.
  •  
26.
  • Galushin, Sergey, et al. (författare)
  • The effect of vessel failure and melt release on risk of containment failure due to ex-vessel steam explosion in Nordic BWR
  • 2019
  • Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE.
  • Konferensbidrag (refereegranskat)abstract
    • Effectiveness of the severe accident management strategy in Nordic BWR reactors depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in the ROAAM+ Framework. The Risk Oriented Accident Analysis Methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR where Melt Ejection Surrogate Model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt Ejection SM is based on the MELCOR code. Modelling of vessel failure and melt release from the vessel in MELCOR code is based on parametric models, allowing a user to select different assumptions and control lower head behavior and melt release. The goal of this work is to evaluate the effect of epistemic uncertain parameters and modelling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR.
  •  
27.
  • Galushin, Sergey, et al. (författare)
  • Treatment of Phenomenological Uncertainties in Level 2 PSA for Nordic BWR Using Risk Oriented Accident Analysis Methodology
  • 2022
  • Ingår i: Probabilistic Safety Assessment and Management, PSAM 2022. - : International Association for Probablistic Safety Assessment and Management (IAPSAM).
  • Konferensbidrag (refereegranskat)abstract
    • A comprehensive and robust assessment of phenomenological uncertainties is a challenge for the current real-life PSA L2 applications, since such uncertainty is majorly driven by physical phenomena and timing of events. Typically, the static PSA models are built on a pre-determined set of scenario parameters to describe the accident progression sequence and use a limited number of simulations in the underlying deterministic analysis to evaluate the consequences. The Risk Oriented Accident Analysis Methodology (ROAAM+) has been developed to enable consistent and comprehensive treatment of both epistemic and aleatory sources of uncertainty in risk quantification. The framework is comprised of a set of deterministic models that simulate different stages of the accident progression, and a probabilistic platform that performs quantification of the uncertainty in conditional containment failure probability. This information is used for enhanced modeling in the PSA-L2 for improved definition of sequences, where information from the ROAAM is used to refine PSA model resolution regarding risk important accident scenario parameters, that can be modelled within the PSA. This work presents an example of application of the dynamic approach in a large-scale PSA model and demonstrate the integration of the ROAAM+ results in the PSA model.
  •  
28.
  • Galushin, Sergey, et al. (författare)
  • Uncertainty analysis of vessel failure mode and melt release in station blackout scenario in Nordic BWR using MELCOR code
  • 2019
  • Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 1944-1956
  • Konferensbidrag (refereegranskat)abstract
    • Nordic Boiling Water Reactors (BWRs) employ containment pressure relief and filtered venting and ex-vessel debris coolability in the deep pool located under the reactor pressure vessel as a severe accident management (SAM) strategy. Effectiveness of this strategy depends on melt release conditions from the vessel which is the major source of uncertainty in risk quantification of containment failure in Nordic BWRs in ROAAM+ Framework. In this work we focus on uncertainty analysis of vessel failure mode and melt release conditions in unmitigated station blackout accident in Nordic BWR. The timing and mode of vessel failure and melt release conditions are predicted by MELCOR code. We demonstrate the effect of MELCOR code uncertain modelling parameters and modelling options on the resultant uncertainty in vessel failure mode and melt release conditions. Results of analysis show that penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, however it does not preclude eventual creep-rupture failure of the vessel lower head. 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. 
  •  
29.
  • Geffray, C., et al. (författare)
  • Verification and validation and uncertainty quantification
  • 2018
  • Ingår i: Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors. - : Elsevier. - 9780081019801 - 9780081019818 ; , s. 383-405
  • Bokkapitel (övrigt vetenskapligt/konstnärligt)abstract
    • In this chapter, an overview of the verification, validation, and uncertainty quantification process is offered. First, the context of the dialog with the safety authorities is explained, and the need for a thorough code validation procedure able to meet the regulatory safety requirements is highlighted. Then, the concept of code verification is introduced, and the main steps are described. The validation process is depicted next. Emphasis is made upon the identification of the physical phenomena of interest and upon the choice of adequate computational tools to capture them. The targeted validity domain of these computational tools and its dependence on available and accurate experimental data are detailed with respect to the issue of scaling. Finally, an overview of selected techniques for uncertainty and sensitivity analysis is provided. 
  •  
30.
  • Grishchenko, Dmitry, et al. (författare)
  • Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR
  • 2017
  • Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery, Inc.
  • Konferensbidrag (refereegranskat)abstract
    • 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. 
  •  
31.
  • Grishchenko, Dmitry, et al. (författare)
  • Risk of containment failure due to ex-vessel steam explosion for Nordic BWRs
  • 2019
  • Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 4032-4038
  • Konferensbidrag (refereegranskat)abstract
    • 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.
  •  
32.
  •  
33.
  • Grishchenko, Dmitry, et al. (författare)
  • Validation of a full model for the analysis of ex-vessel steam explosion in LWRs
  • 2019
  • Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. ; , s. 4568-4574
  • Konferensbidrag (refereegranskat)abstract
    • In a Light Water Reactor (LWR) severe accident, the reactor core can be melted and released from the reactor vessel at ~3000K. In most of reactor designs ex-vessel severe accident mitigation strategy employs a pool of water underneath the reactor vessel. If water pool is deep enough, the melt is expected to be fragmented and quenched and form a coolable debris bed preventing further accident progression. However, there is a possibility that upon contact with volatile coolant thermal energy stored in the hot melt will be converted into mechanical energy of rapidly expanding steam in the process of so called “steam explosion”. Energetics of the steam explosion is contingent upon conditions of melt release, pool characteristics and containment geometry. Containment failure due to the ex-vessel steam explosion can be a factor of risk for the “wet cavity” strategy if fragility limits are close to the expected loads. In order to assess the risk, we develop so called full model (based on TEXAS-V code) for the estimation of the steam explosion loads. To ensure model applicability to a wide range of LWR designs, a number of modifications have been introduced in comparison to previous works. A large database of Full Model solutions is used then for the development of a Surrogate Model based on the Artificial Neural Networks (ANN) to enable extensive sensitivity analysis and uncertainty quantification. The uncertainty in the SM approximation of the FM is considered explicitly in the assessment of failure probability. In this work, we demonstrate an approach to the validation of the Full Model against previous steam explosion experiments using a statistical approach in which a joint distribution of the experimental data is compared to a database of explosion distributions obtained using the full model.
  •  
34.
  • Hernandez, Cuauhtemoc Reale, et al. (författare)
  • Development of a CFD-based model to simulate loss of flow transients in a small lead-cooled reactor
  • 2022
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 392, s. 111773-
  • Tidskriftsartikel (refereegranskat)abstract
    • With the deployment of advanced and small modular reactors (SMRs), it is important to develop the tools to assess their safety. This work presents the different components of a CFD based model for simulating transients in a pool-type small lead cooled reactor. The model encompasses the entire primary circuit with a simplification of the fuel channels, pumps and steam generators. Those parts are modelled through heat and momentum sources (or sinks), similar to the porous medium used in other studies. The CFD solver is coupled with a finite volume solver for fuel pin temperature and a point kinetics solver for neutronics. Free surface is modelled in CFD with multiphase volume of fluid method. The set of methods that is used in this work constitute a novelty for modelling lead cooled reactors. The goal is to have a model that is relatively simple to implement in order to study the effect of some parameters on reactor transients like an unprotected loss of flow. The focus of this study is to describe in detail every individual component of the model, namely the fuel channels, fuel pin temperature, neutronics, coupling strategy, pump and steam generators. In addition, CFD simulations are compared against experimental data from the TALL-3D facility. The purpose of this comparison is to verify that the models and parameters of the CFD software (STAR-CCM+) are capable of reproducing a flow of heavy metal. A future publication will provide the simulation results of an integrated model with all the components.
  •  
35.
  • Jeltsov, Marti, 1985-, et al. (författare)
  • Pre-test analysis of an LBE solidification experiment in TALL-3D
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X.
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper presents the process of development of a solidification test section design for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop of prototypic height). Coolant solidification is a phenomenon of potential safety importance for liquid metal cooled fast reactors (LMFRs). Solidification, e.g. in case of excessive performance of the passive decay heat removal systems, can affect local heat transfer and even lead to partial or complete blockage of the coolant flow paths. This might lead to failure of decay heat removal function. In case of reduced flow circulation, the temperature of the coolant will increase, which might prevent complete blockage of the flow. Prediction of possible outcomes of such scenarios with complex interactions between local physical phenomena of solidification and system scale natural circulation behavior is subject to modelling (epistemic) uncertainty. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support experiment development. The aim of the experimental design is to satisfy requirements stemming from the process of qualification of the model that can be used for addressing the safety-related concerns. In this work we focus on two aspects: (i) design of solidification test section (STS) for investigation of local solidification phenomena of lead-bismuth eutectic (LBE), and (ii) effect of the STS on the system scale behavior of the experimental facility. We discuss selection of the STS characteristics and experimental test matrix using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.
  •  
36.
  • Jeltsov, Marti, 1985-, et al. (författare)
  • Pre-test analysis of an LBE solidification experiment in TALL-3D
  • 2018
  • Ingår i: Nuclear Engineering and Design. - : Elsevier Ltd. - 0029-5493 .- 1872-759X. ; 339, s. 21-38
  • Tidskriftsartikel (refereegranskat)abstract
    • Coolant solidification is a phenomenon of potential safety importance for Liquid Metal Cooled Fast Reactors (LMFRs). Coolant solidification can affect local flow, heat transfer and lead to partial or complete blockage of the coolant flow paths jeopardizing decay heat removal function. It is also possible that reduced flow circulation may increase coolant temperature, counteract solidification and prevent complete blockage of the flow. Complex interactions between local physical phenomena of solidification and system scale natural circulation make modelling of solidification uncertain. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support of experiment development, specifically, design of a solidification test section and a test matrix for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop). The aim of the analysis is experimental design that satisfies requirements stemming from the process of model qualification. We focus on two aspects: (i) design of solidification test section (STS) for investigation of solidification phenomena in lead-bismuth eutectic (LBE), and (ii) effect of the STS pool on the system scale behavior of the TALL-3D facility. Selection of the STS characteristics and experimental test matrix is supported using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes. 
  •  
37.
  • Jeltsov, Marti, 1985-, et al. (författare)
  • Steam generator leakage in lead cooled fast reactors : Modeling of void transport to the core
  • 2018
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 328, s. 255-265
  • Tidskriftsartikel (refereegranskat)abstract
    • Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.
  •  
38.
  • Jeltsov, Marti, 1985- (författare)
  • Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
  • 2018
  • Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
    • Carbon-free nuclear power production can help to relieve the increasing energy demand in the world. New generation reactors with improved safety, sustainability, reliability, economy and security are being developed. Lead-cooled fast reactor (LFR) is one of them.Lead-cooled reactors have many advantages related to the use of liquid metal coolants and pool-type primary system designs. Still, several technological challenges are to be overcome before their successful commercial deployment. The decisions on new designs and licensing of LFRs require use of code simulations, due to lack of operational experience and the fact that full scale experimental investigations are impractical. Code development and validation is necessary in order to reduce uncertainty in predictions for risk assessment and decision support. In order to make the decisions robust, i.e. not sensitive to remaining uncertainty, the process of collecting necessary evidences should iteratively converge. Extensive sensitivity and uncertainty analysis is necessary in order to make the process user-independent.The pool-type design introduces 3D phenomena that require application of advanced modeling tools such as Computational Fluid Dynamics (CFD). The focus of this thesis is on the development and application of CFD modeling and general code validation methodology to the following issues of safety importance for LFR: (i) pool thermal stratification and mixing, (ii) steam generator tube leakage, (iii) seismic sloshing, and (iv) solidification.Thermal mixing and stratification in a pool-type reactor can affect natural circulation stability and thus decay heat removal capability and pose thermal stresses on the reactor structures. Standalone System Thermal-hydraulics (STH) and CFD codes are either inadequate or too computationally expensive for analysis of such phenomena in prototypic conditions. In this work an approach to coupling of STH and CFD is proposed and implemented. For validation of standalone and coupled codes TALL-3D facility (lead-bismuth eutectic (LBE) loop) and test matrix was designed featuring significant two-way thermal-hydraulic feedbacks between the pool-type test section and the rest of the loop.The possibility of core voiding due to bubble transport in case of steam generator tube leakage/rupture (SGTL/R) is a serious safety concern that can be a showstopper for LFR licensing. Voiding of the core increases risks for reactivity insertion accident (RIA) and/or local fuel damage due to deteriorated heat transfer. Bubble transport analysis for the ELSY (European Lead-cooled SYstem) reactor design demonstrated the effect of uncertainty in drag correlation, bubble size distributions and leak location on the core and primary system voiding. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.The effect of seismic isolation (SI) system on the sloshing of the heavy metal coolant and respective mechanical loads on the vessel structures is studied for design basis and beyond design basis earthquakes. It was found that SI shifts the frequencies closer to resonance conditions for sloshing, thus increasing the loads due to slamming of coolant waves onto the vessel internal structures including reactor lid. Possible mitigation measures, such as partitioning baffles have been proposed and their effectiveness was analyzed. Coolant solidification is another serious safety issue in liquid metal cooled reactors that can cause partial or even full blockage of the coolant flow path and thermo-mechanical loads on the primary system structures. Validation of melting/solidification models implemented in CFD is a necessary pre-requisite for its application to risk analysis. In this work, the parametric analysis in support of the design of a solidification test section (STS) in TALL3D facility is carried out. The goals and requirements for the validation experiment and how they are satisfied in the design are discussed in the thesis.
  •  
39.
  •  
40.
  • Jeltsov, Marti, 1985-, et al. (författare)
  • Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 341, s. 306-325
  • Tidskriftsartikel (refereegranskat)abstract
    • Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).
  •  
41.
  • Konovalenko, Alexander, 1975-, et al. (författare)
  • Controllable Generation of a Submillimeter Single Bubble in Molten Metal Using a Low-Pressure Macrosized Cavity
  • 2017
  • Ingår i: Metallurgical and materials transactions. B, process metallurgy and materials processing science. - : SPRINGER. - 1073-5615 .- 1543-1916. ; 48:2, s. 1064-1072
  • Tidskriftsartikel (refereegranskat)abstract
    • We develop a method for generation of a single gas bubble in a pool of molten metal. The method can be useful for applications and research studies where a controllable generation of a single submillimeter bubble in opaque hot liquid is required. The method resolves difficulties with bubble detachment from the orifice, wettability issues, capillary channel and orifice surfaces degradation due to contact with corrosive hot liquid, etc. The macrosized, 5- to 50-mm(3) cavity is drilled in the solid part of the pool. Flushing the cavity with gas, vacuuming it to low pressure, as well as sealing and consequent remelting cause cavity implosion due to a few orders in magnitude pressure difference between the cavity and the molten pool. We experimentally demonstrate a controllable production of single bubbles ranging from a few milliliters down to submillimeter size. The uncertainties in size and bubble release timing are estimated and compared with experimental observations for bubbles ranging within 0.16 to 4 mm in equivalent-volume sphere diameter. Our results are obtained in heavy liquid metals such as Wood's and Lead-Bismuth eutectics at 353 K to 423 K (80 A degrees C to 150 A degrees C).
  •  
42.
  • Kudinov, Pavel, 1972-, et al. (författare)
  • Application of integrated deterministic-probabilistic safety analysis to assessment of severe accident management effectiveness in Nordic BWRs
  • 2016
  • Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
  • Konferensbidrag (refereegranskat)abstract
    • 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.
  •  
43.
  • Kudinov, Pavel, 1972-, et al. (författare)
  • Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness
  • 2019
  • Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. ; , s. 2519-2535
  • Konferensbidrag (refereegranskat)abstract
    • 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.
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44.
  • Kudinov, Pavel, 1972-, et al. (författare)
  • Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials
  • 2017
  • Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 314, s. 182-197
  • Tidskriftsartikel (refereegranskat)abstract
    • Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.
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45.
  • Kumar, Ranjan, et al. (författare)
  • A PSA Level-1 method with repairable components : An application to ASTRID Decay Heat Removal systems
  • 2014
  • Ingår i: Safety and Reliability: Methodology and Applications. - : CRC Press. ; , s. 1611-1617
  • Bokkapitel (övrigt vetenskapligt/konstnärligt)abstract
    • Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I&C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RFT) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP. 
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46.
  • Kööp, Kaspar, 1984- (författare)
  • Application of a system thermal-hydraulics code to development of validation process for coupled STH-CFD codes
  • 2018
  • Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
    • Generation IV reactors are designed to provide sustainable energy generation, minimize waste production and excel in safety. Due to lack of operational experience, ever evolving design and stringent safety requirements, these novel reactors have to rely heavily on simulations.Best estimate one-dimensional (1D) system thermal-hydraulics (STH) codes, originally intended for simulating water-cooled reactor systems with high coolant mass flow rates, are unable to capture complex three-dimensional (3D) phenomena in liquid metal cooled pool-type reactors. Computational fluid dynamics (CFD) codes are capable of resolving the 3D effects, however applying these methods with high resolution for the whole primary system results in prohibiting computational cost.At the same time, there are system components where flow can, with reasonable accuracy, be approximated with 1D models (e.g. core channels, some heat exchangers, etc.). One of the proposed solutions in order to achieve adequate accuracy and affordable computational efficiency in modelling of a Generation IV reactor is to divide the primary system into 1D and 3D regions and apply coupled STH and CFD codes on the respective sub-domains.Successful validation is a prerequisite for application of both, standalone and coupled STH and CFD codes in design and safety analysis of Generation IV systems. In this work we develop and apply different aspects of code validation methodology with an emphasis on (i) STH code analysis in support of validation experiment design (facility and test conditions), (ii) calibration of uncertain code input parameters and validation of standalone STH code, (iii) development of an approach to couple STH and CFD codes.A considerable part of the thesis work is related to the development of a loop-type, 3 leg, liquid metal experimental facility TALL-3D for code validation. Particular focus was on identification of test conditions featuring complex feedbacks between 1D-3D phenomena, which can be challenging for the codes. Standalone STH code (RELAP5) was validated against experimental data. The domain of natural circulation instabilities in TALL-3D operation parameters was discovered using a validated STH code and global optimum search algorithms. Then existence of growing natural circulation oscillations was experimentally confirmed. An international benchmark was initiated in the framework of EU SESAME project based on the obtained experimental data.Simulations were performed to define dimensions and location of a new test section for coolant solidification experiments that would also enhance possibilities for studying natural circulation instabilities in the future tests.An approach to automated input calibration and code validation is developed in order to minimize possible “user effect” in case of multiple uncertain input parameters (UIPs) and system response quantities (SRQs). These methods were applied extensively in the development of RELAP5 input models and identification of the natural circulation instability regions.Domain overlapping approach to coupling of RELAP5 and Star-CCM+ codes was proposed and resulted in considerable improvement of the predictive capabilities in comparison to standalone RELAP5.
  •  
47.
  • Kööp, Kaspar, 1984-, et al. (författare)
  • Automated calibration and validationof RELAP5 input model against TALL-3D facility experimental data
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X.
  • Tidskriftsartikel (refereegranskat)abstract
    • Validation of System Thermal Hydraulics (STH) codes against liquid metal facilities is necessary to increase confidence in designing and licensing of generation IV nuclear power systems. Manual input calibration and tuning against a single set of data can lead to bias in the result of the simulation towards specific system configuration and operation regime.In this work we demonstrate an approach to validation of the RELAP5 code, specifically, applicability of RELAP5 to model complex transients from forced to natural circulation in TALL-3D facility with Lead Bismuth Eutectic (LBE) coolant. We utilize an automated approach to (i) calibration of the input model using different experimental data and (ii) quantification of the modelling uncertainties. The automated approach is intended to reduce the effect of the user on the validation outcomes.Results from the calibrated model are compared against an experiment and uncertainty bounds presented. We discuss the results, provide recommendation to the modelling and provide conclusions on the applicability of the RELAP5 to simulation of different transients.
  •  
48.
  • Li, Hua, et al. (författare)
  • Thermal stratification and mixing in a suppression pool induced by direct steam injection
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 111, s. 487-498
  • Tidskriftsartikel (refereegranskat)abstract
    • An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool. 
  •  
49.
  •  
50.
  • Moreau, V., et al. (författare)
  • Pool CFD modelling : lessons from the SESAME project
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 355
  • Tidskriftsartikel (refereegranskat)abstract
    • The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.
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