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  • 51.
    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.

  • 52.
    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.

  • 53.
    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.

  • 54.
    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.

  • 55.
    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.

  • 56.
    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)
  • 57.
    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.

  • 58. 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.

  • 59.
    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.

12 51 - 59 of 59
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