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  • 51.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    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.
    Lappalainen, J.
    Development of tall-3d test matrix for APROS code validation2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, p. 4562-4575Conference paper (Refereed)
    Abstract [en]

    APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.

  • 52.
    Moreau, V.
    et al.
    CRS4, Sci & Technol Pk Polaris Piscina Manna, I-09010 Pula, Italy..
    Profir, M.
    CRS4, Sci & Technol Pk Polaris Piscina Manna, I-09010 Pula, Italy..
    Alemberti, A.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Frignani, M.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Merli, F.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Belka, M.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Frybort, O.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Melichar, T.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Tarantino, M.
    ENEA FSN ING, I-40033 Camugnano, BO, Italy..
    Franke, S.
    HZDR, Bautzner Landstr 400, D-01328 Dresden, Germany..
    Eckert, S.
    HZDR, Bautzner Landstr 400, D-01328 Dresden, Germany..
    Class, A.
    KIT, Kaiserstr 12, D-76131 Karlsruhe, Germany..
    Yanez, J.
    KIT, Kaiserstr 12, D-76131 Karlsruhe, Germany..
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Roelofs, F.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Zwijsen, K.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Visser, D. C.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Badillo, A.
    PSI, CH-5232 Villigen, Switzerland..
    Niceno, B.
    PSI, CH-5232 Villigen, Switzerland..
    Martelli, D.
    UNIPI, DICI, Largo Lucio Lazzarino,2, I-56122 Pisa, Italy..
    Pool CFD modelling: lessons from the SESAME project2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 355, article id UNSP 110343Article in journal (Refereed)
    Abstract [en]

    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.

  • 53. Papukchiev, A.
    et al.
    Geffray, C.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of the system thermal-hydraulics code ATHlet for the simulation of transient lead-bismuth eutectic flows2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 810-822Conference paper (Refereed)
    Abstract [en]

    Within the European SESAME project, system thermal-hydraulics (STH), computational fluid dynamics (CFD) and coupled 1D-3D thermal-hydraulic simulations were carried out for Generation IV nuclear systems. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH participated in the project with activities related to the development and validation of CFD and coupled CFD-STH codes. The TALL-3D facility, operated by the KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant. A well-instrumented, partially heated test section with cylindrical form is installed in the primary circuit, which is domain of complex 3D flow and heat transfer phenomena. Three different experiments were calculated within a benchmark, organized by KTH: two with coupled programs and one with STH stand-alone code. This paper focuses on the analysis of the observed thermal-hydraulic flow phenomena during the TG03.S310.01 experiment and the comparison between ATHLET predictions and data.

  • 54. Papukchiev, A.
    et al.
    Geffray, C.
    Jeltsov, M.
    Kööp, K.
    Kudinov, P.
    Grischenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Multiscale analysis of forced and natural convection including heat transfer phenomena in the tall-3D experimental facility2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 4, p. 2917-2930Conference paper (Refereed)
    Abstract [en]

    Within the European FP7 project THINS (Thermal Hydraulics of Innovative Nuclear Systems), numerical tools for the simulation of the thermal-hydraulics of next generation rector systems were developed, applied and validated for innovative coolants. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, the Technische Universitaet Muenchen (TUM) and the Royal Institute of Technology (KTH) participated in THINS activities related to the development and validation of computational fluid dynamics (CFD), System Thermal Hydraulics (STH) and coupled STH - CFD codes. High quality measurements from the experiments performed at the TALL-3D facility, operated by KTH, were used to assess the numerical results. This paper summarizes the work accomplished for the validation of the coupled codes ATHLET-ANSYS CFX and RELAP5/STAR CCM+ and highlights the main results achieved for the T01.09 experiment.

  • 55. Papukchiev, A.
    et al.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Importance of conjugate heat transfer modeling in transient CFD simulations2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 615-626Conference paper (Refereed)
    Abstract [en]

    Within the European SESAME project, system thermal-hydraulics (STH), computational fluid dynamics (CFD) and coupled 1D-3D thermal-hydraulic simulations are being carried out for Generation IV nuclear systems. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH participated in the project with activities related to the development and validation of CFD and coupled CFD-STH codes. The TALL-3D facility, operated by the KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant. A well-instrumented, partially heated test section with cylindrical form is installed in the primary circuit. It is the domain of complex 3D flow and heat transfer phenomena. The CFD analyses showed that only the consideration of the fluid domain is not sufficient for correct prediction of the thermal-hydraulic transient dynamics. Therefore, solid structures like test section walls, inner circular plate, test section heater and even the insulation were explicitly included in the CFD model. This allowed the consideration in the simulations of the heat conduction and the thermal inertia of these components. The paper focuses on the analysis of the observed thermal-hydraulic flow phenomena during the TG03.S301.04 experiment and the comparison between ANSYS CFX predictions and data. Moreover, the influence of conjugate heat transfer (CHT) on the numerical results is investigated and discussed.

  • 56. 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.

  • 57.
    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.

  • 58.
    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.

  • 59.
    Phung, Viet-Anh
    et al.
    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.
    Vorobyev, Yury
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Automation of RELAP5 input calibration and code validation using genetic algorithm2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 300, p. 210-221Article in journal (Refereed)
    Abstract [en]

    Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called "user effects". The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the SRQ of primary interest. The ranges of the input parameter were defined based on the experimental data and results of the calibration process. Then GA was used in order to identify combinations of the uncertain input parameters that provide maximum deviation of code prediction results from the experimental data. Such approach provides a conservative estimate of the possible discrepancy between the code result and the experimental data.

  • 60.
    Phung, Viet-Anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Rohde, Martin
    Input Calibration and Validation of RELAP5 Against CIRCUS-IV Single Channel Tests on Natural Circulation Two-Phase Flow Instability2015In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, Vol. 2015, p. 1-14, article id 130741Article in journal (Refereed)
    Abstract [en]

    RELAP5 is a system thermal-hydraulic code that is used to perform safety analysis on nuclear reactors. Since the code is based on steady state, two-phase flow regime maps, there is a concern that RELAP5 may provide significant errors for rapid transient conditions. In this work, the capability of RELAP5 code to predict the oscillatory behavior of a natural circulation driven, two-phase flow at low pressure is investigated. The simulations are compared with a series of experiments that were performed in the CIRCUS-IV facility at the Delft University of Technology. For this purpose, we developed a procedure for calibration of the input and code validation. The procedure employs (i) multiple parameters measured in different regimes, (ii) independent consideration of the subsections of the loop, and (iii) assessment of importance of the uncertain input parameters. We found that predicted system parameters are less sensitive to variations of the uncertain input and boundary conditions in high frequency oscillations regime. It is shown that calculation results overlap experimental values, except for the high frequency oscillations regime where the maximum inlet flow rate was overestimated. This finding agrees with the idea that steady state, two-phase flow regime maps might be one of the possible reasons for the discrepancy in case of rapid transients in two-phase systems.

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  • 61.
    Udalo, Yu. P.
    et al.
    St. Petersburg State Institute of Technology (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Grishchenko, Dmitry
    St. Petersburg State Institute of Technology (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Kulakov, V. V.
    St. Petersburg State Institute of Technology (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Poznyak, I. V.
    St. Petersburg State University of Electrical Engineering, ul. Professora Popova 5, St. Petersburg, 197376 Russia.
    Pechenkov, A. Yu.
    St. Petersburg State University of Electrical Engineering, ul. Professora Popova 5, St. Petersburg, 197376 Russia.
    Phase Separation in Melts of the ZrO2–Al2O3 System2008In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 34, no 5, p. 623-633Article in journal (Refereed)
    Abstract [en]

    The morphology of the quenched and slowly crystallized samples in the ZrO2–Al2O3system isinvestigated in the composition range 25–70 wt % ZrO2. It is revealed that, irrespective of the cooling rate, thesamples contain large baddeleyite (or corundum) crystals, eutectic mixtures, and characteristic regions of intergrown elongated baddeleyite and corundum grains with micron sizes. These regions have the same phase composition at any initial ratio between zirconium and aluminum oxides and at any cooling rates of the melt. Ahypothesis is put forward that these regions are products of the decomposition of ZrO2· 2Al2O3associates.

  • 62.
    Udalov, Yu. P.
    et al.
    St. Petersburg State Technological Institute (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Grishchenko, Dmitry
    St. Petersburg State Institute of Technology (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Liquation in melts of oxide systems2006In: Collected papers of Russian Scientific and Technical conference “Electrothermy – 2006”, 2006, p. 29-51Conference paper (Refereed)
  • 63.
    Udalov, Yu. P.
    et al.
    St. Petersburg State Technological Institute (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Grishchenko, Dmitry
    St. Petersburg State Institute of Technology (Technical University), Moskovskii pr. 26, St. Petersburg, 198013 Russia.
    Petrov, Yu. B.
    St. Petersburg State University of Electrical Engineering, ul. Professora Popova 5, St. Petersburg, 197376 Russia.
    Poznyak, I. V.
    St. Petersburg State University of Electrical Engineering, ul. Professora Popova 5, St. Petersburg, 197376 Russia.
    Pechenkov, A. Yu.
    St. Petersburg State University of Electrical Engineering, ul. Professora Popova 5, St. Petersburg, 197376 Russia.
    Monotectic Crystallization of Melts in the ZrO2–Al2O3 System2006In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 32, no 4, p. 479-485Article in journal (Refereed)
    Abstract [en]

    —The morphology of the crystalline phases prepared at different cooling rates, temperatures, andcompositions of melts in the ZrO2–Al2O3system is investigated. It is established that both the quenched andslowly crystallized samples containing 35–70 wt % ZrO2have a submicron structure. Outside this concentrationrange, the ingots have a zonal structure: the peripheral region is formed by large baddeleyite crystals (at a ZrO2content higher than 70 wt %) or corundum crystals (at a ZrO2content lower than 30 wt %). This character ofthe crystallization confirms the presence of the phase separation (liquid immiscibility) region in melts of thissystem. A variant is proposed for the phase diagram of the system under investigation.

  • 64. Udalov, Yurii
    et al.
    Pozniak, I.V.
    Pechenkov, A.Yu.
    Sazavsky, P.
    Kiseleva, M.
    Schrank, I
    Pospekhova, J
    Piluzo, P.
    Grishchenko, Dmitry
    Coordination Nature of Phase Separation in Oxide Melts2013In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 39, no 4, p. 431-443Article in journal (Refereed)
    Abstract [en]

    The behavior of uranium oxide-silica, zirconia-alumina, zirconia-iron oxide, uranium oxideiron oxide, zirconia-alumina-iron oxide, and uranium oxide-iron oxide-zirconia melts was experimentally investigated in the air. The existence of two-phase fluids in these systems was confirmed. It is proposed that the reason for the phase separation is the formation of complexes with the general formula x[M3+O8/2] [Me4+O8/2]. The influence of a complex concentration on the density and surface tension of melts in the ZrO2-Al2O3 system was demonstrated.

  • 65.
    Wang, Xicheng
    et al.
    KTH.
    Gallego-Marcos, Ignacio
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Post-test calibration of the Effective Momentum Source (EMS) model for steam injection through multi-hole spargers.2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6176-6189Conference paper (Refereed)
    Abstract [en]

    Steam condensation in a large pool is often used in light water reactors to prevent containment overpressure. In boiling water reactors, steam from the primary system can be released into a pressure suppression pool (PSP) in normal operation and during accidents through multi-hole spargers to control the pressure in the reactor vessel. Steam injection into the pool can lead to the development of thermal stratification that affects (i) pressure suppression capacity of the pool, (ii) operation of the safety systems that use PSP as a source of water (e.g. emergency core cooling system and containment spray). Modeling of direct contact condensation of steam presents a challenge for contemporary codes. Therefore, Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. EMS defines the time-averaged effect of steam injection into the pool in terms of a momentum source. For multi-hole spargers, the momentum source requires models for (i) momentum induced by multi-holes steam injection, (ii) direction (vertical angle) of the induced momentum, and profile of velocity in (iii) vertical and (iv) azimuthal directions. Previous works on EMS model validation and sensitivity study against PPOOLEX and HYMERES PANDA pool tests suggest the importance of all these factors for accurate prediction of the pool mixing behaviour. All these parameters, except the velocity profile in the azimuthal direction, were measured in PANDA facility and in Separate Effect Facility (SEF) at Lappeenranta Institute of Technology. The goal of this work is to develop a model for the azimuthal profile of radial velocity (APV) of water induced by steam injection through multi-hole spargers in a pressure suppression pool of a Nordic Boiling Water Reactor (BWR). In previous work, it was assumed that the APV is the same as the radial velocity profile in vertical cross section (which was measured in PANDA experiments using PIV) and can be described by axisymmetric jet expansion model. In this paper, APV is defined as a separate model with own closure for the jet diffusion rate. The effect of the steam mass flow rate is taken into account in the APV and respective jet expansion factor according to the experimental observations. Finally, we compare the pool temperature evolution in the experiment and simulations with the EMS model.

  • 66.
    Wang, Xicheng
    et al.
    KTH.
    Gallego-Marcos, Ignacio
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Pre-test analysis for HYMERES-2 PANDA tests series for steam injection into pool through spargers2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6190-6203Conference paper (Refereed)
    Abstract [en]

    Steam condensation in a pool of water is often used in light water reactors. Steam injection provides sources of heat and momentum, which can lead to the development of thermal stratification or mixing of the pool. Modelling of the direct steam condensation is computationally challenging, especially considering the complex geometry of the pool and spargers and duration of transients. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed and implemented to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. These models are utilized in this work and are subject for further development. The goal of this work is to provide pre-test analysis to support the design and selection of the test conditions for the Pressure Suppression Pool (PSP) test series (OECD/HYMERES-2). This test series is aiming to provide data for the development and validation of the EMS model predictive capabilities for the PSP phenomena. Specifically, it was proposed to extend the experimental database with regards to the sparger design, pool depth and depth of sparger submergence. In order to maximize the value of obtained data for the model development and validation, the pre-test analysis of HYMERES-2 is carried out. Compared to the HYMERES-1 test series, the sparger elevation above pool bottom will be increased in order to study the effect of the distance between the sparger head and thermocline interface on the rate of erosion of the stratified layer. The Particle Image Velocimetry (PIV) will be used for the measurement of the azimuthal profile of flow radial velocity around the sparger. A large number of thermocouples are provided in the vertical direction to capture the transient location of the thermocline. Selection of specific values for the sparger elevation, location of the PIV window, number of thermocouples, and steam injection conditions are based on the analysis provided in this work. In the pre-test analysis we use the previous EHS/EMS models which show a good agreement with the HYMERES-1 test series.

  • 67.
    Zabiego, Magali
    et al.
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Brayer, Claude
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Grishchenko, Dmitry
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Dajon, Jean-Baptiste
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Fouquart, Pascal
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Bullado, Yves
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Compagnon, Frédéric
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France.
    Correggio, Patricia
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France .
    Haque, Jean-François
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, France .
    Piluso, Pascal
    CEA, DEN, STRI, LMA, F-13108 Saint-Paul-Lez-Durance, Franc.
    The KROTOS KFC and SERENA/KS1 tests: experimental results and MC3D calculations2010In: 7th International Conference on Multiphase Flow ICMF 2010, 2010Conference paper (Refereed)
    Abstract [en]

    During a hypothetical severe accident in a Pressurized Light Water Reactor, the hot molten material mixture (corium) issuingfrom the degraded reactor core may generate a steam explosion if it comes in contact with water. Such an explosion maydamage the structures and threaten the reactor integrity. The SERENA program is an international OECD project which aims athelping the understanding of the interaction mechanisms between the different components of the system (the hot molten pool,the liquid water, the generated fragments and steam) by providing experimental data and calculation analysis.In the KROTOS facility, energetic steam explosions can be triggered and studied using prototypic corium. CEA takes part in theSERENA program by performing experimental tests in the KROTOS facility designed to allow direct visual observations ofmelt injection into a water tank and mixing conditions (Huhtiniemi 2001). Still in the SERENA frame, pre- and post-testanalysis are also carried out by CEA with the MC3D software. MC3D is developed by IRSN, it is a thermal-hydraulicmultiphase flow code mainly dedicated to ex-vessel and in-vessel Fuel Coolant Interactions (FCI) studies (Meignen 2005).The purpose of this paper is to present the KROTOS experimental setup and instrumentation, expose the experimental resultsand a MC3D analysis of the recent KROTOS tests. The experimental results are presented for the KFC scoping test of theKROTOS program for which the results are not restricted to the SERENA partners. On the contrary, the KROTOS KS1 resultsbeing only available to the SERENA partners, the pre-mixing and explosion MC3D post-KS1 test calculations are proposed butwithout any scale. The calculation hypothesis leading to a good agreement with the experimental results (in terms of jetprogression and void fraction in the system) are discussed.

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