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  • 151.
    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.

  • 152.
    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.

  • 153.
    Phung, Viet-Anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Rohde, Martin
    Delft University of Technology.
    Simulation of Two-Phase Flow Instability in CIRCUS Facility Using RELAP52008In: Transactions of the American Nuclear Society: Volume 99, American Nuclear Society, 2008, p. 813-814Conference paper (Refereed)
  • 154.
    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.
    Identification of Two-Phase Flow Regimes in Unstable Natural Circulation Using TRACE and RELAP52012In: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012, p. N9P0062-Conference paper (Refereed)
    Abstract [en]

    The paper is concerned with validation of system thermal hydraulic (STH) codes TRACE, RELAP5 and identification of flow regimes against experimental data from low pressure CIRCUS-IV facility (Delft University of Technology, The Netherlands) on two-phase natural circulation flow instability. In previous work [1], RELAP5 was validated against the CIRCUS data. The study showed that RELAP5 can predict relatively well the instability in the system. Later, the instantaneous two-phase flow regime identified in simulations was compared with the flow pattern recorded by high-speed camera. It suggested that the flow regime was often misidentified by the code. In this paper, we focus on the effect of flow regime treatment on two-phase flow instability prediction. Both codes are implicitly using concept of flow regime map and correlations obtained for steady state flow conditions. RELAP5 and TRACE have quite different approaches to the treatment of two-phase flow regimes. While RELAP5 has more elaborate flow regime map, TRACE is using an interpolation between bubbly-slug and annular mist flow regimes to simulate all intermediate regimes. Analysis of the codes prediction results sheds new light on the importance of the flow regime in STH codes. We also provide discussion on sensitivity of the predicted unsteady flow to parameters of flow regime treatment.

  • 155.
    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.
    Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes2015In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 296317Article in journal (Refereed)
    Abstract [en]

    In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii) assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified.

  • 156.
    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.
    Relaxation Time Concept for Flow Regime Transition in Two-Phase Flow Simulations2009In: Transactions of the American Nuclear Society, American Nuclear Society, 2009, p. 868-869Conference paper (Refereed)
  • 157.
    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.

  • 158.
    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.
    Rohde, Martin
    Delft University of Technology.
    Validation of RELAP5 with Sensitivity Analysis for Uncertainty Assessment for Natural Circulation Two-Phase Flow Instability2009In: Proceedings of the 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), 2009, p. 1-18Conference paper (Refereed)
  • 159. Pohlner, G.
    et al.
    Buck, M.
    Meignen, R.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Polidoro, F.
    Takasuo, E.
    Analyses on ex-vessel debris formation and coolability in SARNET frame2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 50-57Article in journal (Refereed)
    Abstract [en]

    The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the view on possible key aspects of future activities.

  • 160.
    Prykhodko, O.A.
    et al.
    Dnepropetrovsk State University.
    Kudinov, Pavel
    Dnepropetrovsk State University.
    Численное моделирование нестационарного взаимодействия подвижных решеток профилей компрессора и трансзвуковых течений с фазовыми переходами2004In: Aerodynamics: problems and perspectives. Kharkov: National Academy of Ukraine Kharkov Institute of Aviation, p. 93-111Article in journal (Refereed)
  • 161. Prykhodko, O.A.
    et al.
    Kudinov, Pavel
    Pismenny, V.I.
    Menaylov, A.V.
    Numerical simulation of transonic gas-vapor flows with condensation2004In: Proceedings of V Minsk International Forum on Heat and Mass Transfer, Institute of Heat and Mass Transfer A.V. Lykova, 2004Conference paper (Refereed)
  • 162.
    Torregrosa, Claudio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled 3D Thermo-mechanical Analysis of a Nordic BWR Vessel Failure and Timing2013In: Proceedings 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Conference paper (Refereed)
  • 163. Tran, C. -T
    et al.
    Nguyen, V. -H
    Tahara, M.
    Kojima, Y.
    Hamazaki, R.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A study on transient heat transfer of the EU-ABWR external core catcher using the Phase-change Effective Convectivity Model2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, American Nuclear Society, 2015, p. 6821-6834Conference paper (Refereed)
    Abstract [en]

    In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation HI and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and procedures are evaluated based on results of PECM simulation and severe accident analysis.

  • 164.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A Synergistic use of CFD, Experiments and Effective Convectivity Model to Reduce Uncertainty in BWR Severe Accident Analysis2010Conference paper (Refereed)
  • 165.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Local Heat Transfer From The Corium Melt Pool to the BWR Vessel Wall2011In: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Conference paper (Refereed)
  • 166.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head2009In: International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009, Atomic Energy Society of Japan , 2009, Vol. 2, p. 1523-1537Conference paper (Refereed)
    Abstract [en]

    The paper is concerned with development of models for assessment of Control Rod Guide Tube (CRGT) cooling efficiency in Severe Accident Management (SAM) for a Boiling Water Reactor (BWR). In case of core melt relocation under a certain accident condition, there is a potential of stratified (with a metal layer atop) melt pool formation in the lower plenum. For simulations of molten metal layer heat transfer we are developing the Effective Convectivity Model (ECM) and Phase-change ECM (PECM). The models are based on the concept of effective convectivity previously developed for simulations of decay-heated melt pool heat transfer. The PECM platform takes into account mushy zone convection heat transfer and compositional convection that enables simulations of non-eutectic binary mixture solidification and melting. The ECM and PECM are validated against various heat transfer experiments for both eutectic and non-eutectic mixtures, and benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is applied to heat transfer simulation of a stratified heterogeneous debris pool in the presence of CRGT cooling. The PECM simulation results show no focusing effect in the metal layer on top of a debris pool formed in the BWR lower plenum and apparent efficacy of the CRGT cooling which can be served as an effective SAM measure to protect the vessel wall from thermal attacks and mitigate the consequences of a severe accident.

  • 167.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head2013In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, Vol. 2013, p. 231501-Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with the development of approaches for assessment of core debris heat transfer and Control Rod Guide Tube (CRGT) cooling effectiveness in case of a Boiling Water Reactor (BWR) severe accident. We consider a hypothetical scenario with stratified (metal layer atop) melt pool in the lower plenum. Effective Convectivity Model (ECM) and Phase-Change ECM (PECM) are developed for the modeling of molten metal layer heat transfer. The PECM model takes into account reduced convection heat transfer in mushy zone and compositional convection that enables simulations of noneutectic binary mixture solidification and melting. The ECM and PECM are (i) validated against relevant experiments for both eutectic and noneutectic mixtures and (ii) benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is then applied to the analysis of heat transfer in a stratified heterogeneous debris pool taking into account CRGT cooling. The PECM simulation results show apparent efficacy of the CRGT cooling which can be utilized as Severe Accident Management (SAM) measure to protect the vessel wall from focusing effect caused by metallic layer.

  • 168.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head2010In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, no 9, p. 2148-2159Article in journal (Refereed)
    Abstract [en]

    This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.

  • 169.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An approach to numerical simulation and analysis of molten corium coolability in a BWR lower head2008Conference paper (Refereed)
  • 170.
    Tran, Chi Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A Study on the Integral Effect of Corium Material Properties on Melt Pool Heat Transfer in a Boiling Water Reactor2011In: Proceedings 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 2011Conference paper (Refereed)
  • 171.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Hua
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Puustinen, Markku
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Generalization of experimental data on amplitude and frequency of oscillations induced by steam injection into a subcooled pool2015In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 295, p. 155-161Article in journal (Refereed)
    Abstract [en]

    Steam venting and condensation into a subcooled pool of water through a blowdown pipe can undergo a phenomenon called chugging, which is an oscillation of the steam water interface inside the blowdown pipe. The momentum that is generated by the oscillations is directly proportional to the oscillations' amplitude and frequency, according to the synthetic jet theory. Higher momentum can enhance pool mixing and positively affect the pool's pressure suppression capacity by reducing thermal stratification. In this paper, we present a generalization of available experimental data on the amplitude and frequency of oscillations during chugging. We use experimental data obtained in different facilities at different scales to suggest a scaling approach for non-dimensional amplitude and frequency of the oscillations. We demonstrate that the Froude number Fr (which relates the inertial forces to gravitational forces) can be used as a scaling criterion in this case. The amplitude has maximum at Fr approximate to 2.8. There is also a strong dependence of the amplitude on temperature; the lower the bulk temperature is the higher the scaled amplitude. A known analytical theory can only capture the decreasing trend in amplitude for Fr > 2.8 and fails to capture the increasing trend and the temperature dependence. Similarly, there is a minimum of the non-dimensional frequency at Fr approximate to 6. A strong dependence on temperature is also observed for Fr> 6; the lower the bulk temperature is the higher the scaled frequency. The known analytical theory is able to capture qualitatively the general trend in frequency but not the dependence on temperature.

  • 172.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A Computational Study On Instrumentation Guide Tube Failure During a Severe Accident in Boiling Water Reactor2011In: The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 2011Conference paper (Refereed)
  • 173.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of Instrumentation Guide Tube Failure in a BWR Lower Head2012In: Proc. 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Conference paper (Refereed)
  • 174.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Assessment with Coupled Thermo-Mechanical Creep Analysis of Combined CRGT and External Vessel Cooling Efficiency for a BWR2011In: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Conference paper (Refereed)
  • 175.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head2012In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 249, p. 146-153Article in journal (Refereed)
    Abstract [en]

    In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a 'ballooning' of the vessel bottom and (ii) a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel cooling was able to suppress the creep and subsequently prevent vessel failure for the considered pool depths.

  • 176.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi-Thanh
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effect Of CRGT Cooling On Modes Of Global Vessel Failure Of A BWR Lower Head2012In: Proceedings Of The 20th International Conference On Nuclear Engineering And The ASME 2012 Power Conference - 2012, Vol2, 2012, p. 467-477Conference paper (Refereed)
    Abstract [en]

    An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) 'ballooning' of the vessel bottom for smaller pools, and (b) 'localized creep' concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

  • 177. Vorobyev, Y.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development and Application of a Genetic Algorithm Based Dynamic PRA Methodology to Plant Vulnerability Search2011In: International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011, 2011, p. 559-573Conference paper (Refereed)
    Abstract [en]

    The paper describes recent achievements in development and application of the Dynamic Probabilistic Risk Analysis (DPRA) methodology based on the Genetic Algorithm (GA). The aim of the GA-DPRA approach is to enable identification of safety vulnerabilities and quantification of accident risks related to operation of nuclear power plants (NPP). The approach combines a system code as a deterministic model of the plant and a GA search engine for the exploration of the plant scenarios space. A point in this space represents a scenario (transient) which is defined by unique combination of initial plant state and time dependent sequence of changes in the plant state parameters implemented in the system code input. The GA-DPRA is used to address two main types of safety analysis problems: (i) identification of a "worst case" scenario with most severe violation of safety limits (failure of safety barriers); (ii) identification of "failure domains" (subdomains in the space of plant scenarios where at least one of the safety limits (barriers) is violated). Safety critical parameters (safety limits) are used by GA as fitness functions to guide selection of the system code input parameters in process of the global optimum search. The GA controls selection of system code input parameters within predefined diapasons and time windows. Unlike "brute force" approaches or Monte Carlo type methods the GA-DPRA is much less demanding to computational resources due to intelligent and adaptive resolution in the exploration of the plant scenarios space. Stochastic properties of GA and Importance Sampling technique are applied to estimate probabilistic characteristics of the identified vulnerabilities. Solutions of benchmark problems and comparison with other methods are discussed in the paper.

  • 178. Vorobyev, Y.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of methodology for identification of failure domains with GA-DPSA2012In: 11th International Probabilistic Safety Assessment and Management Conference and the Annual European Safety and Reliability Conference 2012, PSAM11 ESREL 2012: Volume 3, 2012, p. 2480-2489Conference paper (Refereed)
    Abstract [en]

    Methods of PSA/PRA play important role in understanding of threats to Nuclear Power Plants (NPPs) safety. However, static logic of PSA has difficulties in considering the dynamic nature of physical processes and their interaction with stochastic events. Different Dynamic PSA (DPSA) methods have been proposed to resolve the influence of timing and order of events in safety analysis of NPP. In this work we discuss a DPSA approach which employs global optimum search methods (particularly genetic algorithm (GA)) for the exploration of the uncertainty space (the space of plant accident scenarios and uncertain parameters) and a system code as a deterministic model of the plant. The GA is used to generate the system code input for probing the uncertainty space. Safety important parameters (e.g. peak cladding temperature etc.) are used by GA as objective functions (which are also often called fitness functions in GA) to guide selection of the system code input to find conditions at which safety limits are exceeded (failure domains (FD) in the uncertainty space). The biggest challenges in the problem of FD identification are (i) difficulties caused by enormous dimensionality of the space, (ii) large variations in sensitivity of the fitness function to different input parameters, (iii) significant cross correlations between input parameters, and (iv) non-monotonic behavior of the fitness function in the whole uncertainty space. In this paper we report most recent developments of the GA-DPSA methodology. Specifically we investigate the influence of the selection of GA internal parameters on the efficiency of failure domain identification. A method for probability estimation based on the neural networks is discussed. A test case for DPSA methods is proposed based on the LOCA scenario for a PWR model distributed along with the RELAP5 code. Presented test case reveals intricate dynamic interactions in different accident scenarios despite relative simplicity of the model.

  • 179. Vorobyev, Yu.B.,
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Nhat, T.V.K.
    Application of information technologies (genetic algorithms, neural networks, parallel calculations) in safety analysis of Nuclear Power Plants2014In: Proceedings of the Institute for System Programming, ISSN 2220-6426, Vol. 26, no 2, p. 137-158Article in journal (Refereed)
    Abstract [en]

    This paper investigates important issues in three types of safety assessment methodologies commonly applied for Nuclear Power Plants (NPP). These methodologies are i) dynamic probabilistic safety assessment (DPSA) where application of genetic algorithm (GA) is shown to improve the efficiency of the analysis, ii) deterministic safety assessment (DSA) with meta model representation of the system using pre-performed computational fluid dynamics (CFD) code and iii) vulnerability search (e.g. identification of accident scenarios in an NPP) with application of neural network (NN). The use of advanced computational tools and methods such as genetic algorithms, neural networks and parallel computations improve the efficiency of safety analysis. To achieve the best effect, these advanced technologies are to be integrated with existing classical methods of safety analysis of the NPP.

  • 180. Yakush, S. E.
    et al.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effect of particle spreading on coolability of Ex-Vessel debris BED2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, American Nuclear Society, 2015, p. 1210-1222Conference paper (Refereed)
    Abstract [en]

    Debris bed formation and coolability are studied by DECOSIM code. Main physical mechanisms affecting dispersed particle spreading in the course of melt-water interaction are considered, and their relevance to the formation of porous debris bed in various melt ejection modes is discussed. Numerical simulations of gradually growing and instantly formed debris beds are performed by DECOSIM code. Also, coupled simulations are carried out in which all mechanisms are taken into account simultaneously. It is shown that particle spreading limits the height of debris bed. Also, it is obtained that in some parameter ranges even if local dryout occurs, further particle spreading can render the debris bed coolable, resulting in its reflooding and quenching of the material.

  • 181. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Melt agglomeration influence on ex-vessel debris bed coolability2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

    A deep pool of water below reactor vessel is employed in some light water reactors in order to arrest the core melt progression in case of severe accident. The melt is expected to fragment and quench in the pool. Coolability of a porous debris bed in a water pool is considered in this work, with emphasis on the effect of agglomerates formed due to incomplete fragmentation of a melt jet, or incomplete freezing of melt particles reaching the debris bed. Agglomerates block the escape paths for vapor generated in the debris bed, increasing resistance to coolant flow and facilitating occurrence of dry zones. Numerical simulations are carried out by DECOSIM multiphase code. Two main types of debris bed inhomogeneity due to agglomeration are considered: (i) solid impermeable “cake” on the debris bed top, and (ii) distributed low-permeability low-porosity zone with properties determined according to prediction of an agglomeration model. In the presence of either type of flow obstruction, dryout occurs at significantly lower decay heat power than in the case of homogeneous non- agglomerated debris. There is a critical agglomeration fraction above which a large dry zone develops in the debris bed. Debris temperature escalates in the dry zone leading to subsequent remelting of the material. On the contrary, below the critical agglomeration fraction, the dry zone temperature is stabilized by vapor cooling. Implications of the obtained results for assessment of severe accident risks are discussed.

  • 182. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the evaluation of dryout conditions for a heat-releasing porous bed in a water pool2019In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, p. 895-905Article in journal (Refereed)
    Abstract [en]

    This work is motivated by the problem of restraining temperature escalation inside a porous heat-releasing media submerged in a pool of liquid coolant. When coolant temperature reaches saturation, boiling begins in the bulk of the porous bed, with void generation rate determined by the heating power. Amount of void determines hydrostatic pressure difference that drives natural circulation of two-phase flow through the porous material. At a certain critical value of the heat release rate, the driving head cannot overcome drag of the two phase porous media flow, which results in complete evaporation of coolant in some zone. Temperature of material in the dry zone increases significantly due to deterioration of heat exchange with single phase vapor flow in comparison with boiling flow heat transfer. The paper considers the problem of determining the critical conditions for onset of dryout in a heat-releasing porous bed of an arbitrary shape. The well-known one-dimensional problem for a flat top-flooded bed is revisited, and the functional form of the dryout boundary (expressed as the dryout heat flux, DHF) is derived using non-dimensional parameters. Asymptotic behavior of the solution is analyzed, and, by the method of asymptotic interpolation, a surrogate model is proposed consisting of three single-argument, non-dimensional functions. It is shown that such a model provides acceptable accuracy even in the cases where complete similarity of solutions is not achieved. The results obtained provide important insights into the physics of the problem, reduce the number of free parameters, and enable fast evaluation of dryout conditions without the need of numerical solution of algebraic equations involved in the exact formulation. The ultimate goal of the surrogate model development, i.e. its application to multidimensional configurations, is discussed.

  • 183. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Transient phenomena of ex-vessel debris bed formation in an LWR severe accident2009In: Transactions of the American Nuclear Society, 2009, p. 546-547Conference paper (Refereed)
  • 184. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Lubchenko, N. T.
    Coolability of heat-releasing debris bed. Part 1: Sensitivity analysis and model calibration2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 52, no SI, p. 59-71Article in journal (Refereed)
    Abstract [en]

    Coolability of heat-releasing debris bed is an important issue in the severe accident analysis and management. Traditionally, theoretical studies of top or bottom-fed debris bed coolability have been focused on obtaining a "best estimate" value for the Dryout Heat Flux (DHF) as a function of debris bed parameters (mean particle diameter and porosity). However, an important question for safety analysis is the quantification of uncertainties inherent in the problem. In this paper, a one-dimensional coolability problem is considered, with the aim of analyzing the influence of aleatory uncertainties in input physical parameters and modeling (epistemic) uncertainties on the prediction of DHF. Global sensitivity analysis is applied to rank the aleatory and epistemic parameters according to their effects on DHF and average pressure drop. The most influential model parameters are then calibrated to achieve the best fit to experimental data available. On the one hand, we demonstrate that model calibration is instrumental in achieving considerable improvement of quantitative agreement between the experimental and simulation data. On the other hand, experience of model calibration also suggested that (i) optimization of model parameters with respect to available experimental data on DHF is an ill-posed problem, and (ii) model calibration with respect to one-dimensional pressure drop experiments does not automatically improve the prediction of DHF and in some cases can even worsen it. Based on these insights, one can speculate that further analytical and experimental efforts are necessary to establish a better consistency between model form and experimental data on pressure drop and DHF.

  • 185. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Lubchenko, N. T.
    Coolability of heat-releasing debris bed. Part 2: Uncertainty of dryout heat flux2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 52, no SI, p. 72-79Article in journal (Refereed)
    Abstract [en]

    One-dimensional coolability problem for a flat homogeneous heat-releasing debris bed is considered, with the critical conditions for steady-state cooling characterized by the Dryout Heat Flux (DHF). DHF is determined for top-fed and bottom-fed debris beds from several two-phase models differing by the treatment of porous and interphase drag. Aleatory uncertainties due to randomness of the debris bed formation scenario and respective physical parameters (particle diameter, porosity) are quantified. The influence of ranges and distributions of input parameters on the uncertainty in the DHF are studied. Major contributors to the DHF uncertainty are identified. Influence of model uncertainty on the prediction of the lower boundary for DHF is discussed.

  • 186. Yakush, S. E.
    et al.
    Kudinov, Pavel Yu
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A model for prediction of maximum post-Dryout temperature in decay-heated debris BED2014In: International Conference on Nuclear Engineering, Proceedings, ICONE, 2014, Vol. 2BConference paper (Refereed)
    Abstract [en]

    Several designs of light water reactors consider melt fragmentation and cooling of corium debris bed in a deep pool as important part of their severe accident management strategies. Traditional approach to assessment of debris coolability is based on the bed dryout criterion. However, this is the most conservative criterion which doesn’t take into account possibility of debris temperature stabilization in steam cooling regime. In this work numerical simulations of cooling of a decay heat-releasing porous debris bed in a water pool are carried out for the conditions where local dryout of porous material occurs. It is shown that the temperature of solid material in the dry zone can be stabilized if sufficient vapor flow is generated in the wetted part of the debris bed beneath the dry zone. A simple one-dimensional model which connects the maximum temperature and the relative size of the dry zone is proposed and verified against the numerical simulations with DECOSIM code for different shapes of the debris beds relevant to severe accident conditions in a Nordic type boiling water reactor (BWR). On the basis of this model, a criterion is obtained which defines the critical relative height of the dry zone corresponding to specific temperature of debris material which can be considered as a safety limit (e.g. start of zirconium oxidation, remelting of metallic debris or oxidic corium, etc.). The criterion allows one to evaluate the safety margins and degree of conservatism introduced by the dryoutbased approach to assessment of debris coolability.

  • 187. Yakush, S. E.
    et al.
    Lubchenko, N. T.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Risk-informed approach to debris bed coolability issue2012In: Proceedings Of The 20th International Conference On Nuclear Engineering And The ASME 2012 Power Conference - 2012, Vol 2, ASME Press, 2012, no 1, p. 531-543Conference paper (Refereed)
    Abstract [en]

    Coolability of an ex-vessel debris bed in severe accident conditions is considered from the risk perspective. The concept of "load versus capacity" is employed to quantify the probability of failure (local dryout). Possible choices of "load" and "capacity" in terms of heat fluxes, thermal power or melt mass are discussed. Results of Monte Carlo simulations of distribution functions for the local heat flux and the dryout heat flux at the debris bed top point (defined as the extensions of one-dimensional counterparts) are presented. A surrogate model for the dryout heat flux is developed by the generalization of two-dimensional simulation results. Dryout probabilities are obtained under the conservative assumptions (neglecting the coolability improvement due to side ingress of water into a non-flat debris bed), and from the surrogate model. Outlook is given for the prospective development of the risk-informed approach to debris bed coolability in the context of comprehensive severe accident risk analysis.

  • 188.
    Yakush, S.
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences, Moscow, Russia.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Simulation of Ex-Vessel Debris Bed Formation and Coolability in a LWR Severe Accident2009In: Proceedings of ISAMM 2009: Implementation of severe accident management measures, Paul Scherrer Institute , 2009Conference paper (Refereed)
  • 189.
    Yakush, S.
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences, Moscow, Russia.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Transient Phenomena of Ex-vessel Debris Bed Formation in a LWR Severe Accident2009In: American Nuclear Society Transactions 2009, American Nuclear Society, 2009, p. 546-547Conference paper (Refereed)
  • 190. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Modeling of Two-Phase Natural Convection Flows in a Water Pool with a Decay-Heated Debris Bed2008In: Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP 2008), 2008Conference paper (Refereed)
    Abstract [en]

    Coolability of a debris bed in a water pool is studied numerically with the emphasis on the effects of natural convection flows driven by vapor release in the porous bed. A multifluid computer code DECOSIM is developed and applied to the simulation of liquid-vapor flows in the saturated conditions relevant to the long transient of the debris bed formation and coolability problems. Numerical calculations carried out reveal the influence of the debris bed shape on the structure of natural circulation flows in the water pool, as well as on the void fraction distributions inside the debris bed. The effect of large scale flow structures in the pool on transport of debris particles is demonstrated. It is shown that the vortical flow can capture the smaller particle which may result in considerable spreading of the sedimented debris over the pool floor.

  • 191. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Multiscale Simulations of Self-organization Phenomena in the Formation and Coolability of Corium Debris Bed2009In: Proc. The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), 2009Conference paper (Refereed)
  • 192. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Lubchenko, N.T.
    Sensitivity and Uncertainty Analysis of Debris Bed Coolability2011In: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Conference paper (Refereed)
    Abstract [en]

    Theoretical studies of top-fed debris bed coolability available so far have been focused on obtaining the Dryout Heat Flux (DHF) as a function of debris bed parameters (mean particle diameter and porosity). In this paper, uncertainty analysis is carried out to quantify the influence of different factors on DHF. Global sensitivity analysis is applied to rank the drag model parameters according to their effects on DHF and average pressure drop (epistemic uncertainty). The most influential model parameters are then optimized to achieve the best fit to experimental data available. Finally, aleatory uncertainties due to randomness of the debris bed format

  • 193.
    Yakush, Sergey
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effects of Water Pool Subcooling on the Debris Bed Spreading by Coolant Flow2011In: ICAPP2011: Proceedings of the 2011 international congress on advances in nuclear power plants, American Nuclear Society, 2011Conference paper (Refereed)
    Abstract [en]

    Deep water pool in the reactor pit is considered an effective means of arrest and long term cooling thecorium released from the reactor pressure vessel in the case of a hypothetical severe accident with lightwater reactors. Corium melt interaction with water is expected to result in the material fragmentation andformation of a porous debris bed on the containment basemat. From the safety point of view, it isessential to quantify the conditions for debris bed coolability without corium remelting due to the internalheat release. Previous research results have clearly demonstrated that, given the total mass of the material,the conditions for the occurrence of local dryout are contingent on the particle diameters, porosity, as wellas on the debris bed shape. In particular, a tall heap-shaped debris bed would be more prone to dryoutthan that uniformly spread over the pool basemat.For the BWR designs adopted in Sweden and Finland, one of the severe accident scenarios to beconsidered is the gradual melt release, for which the molten corium is released over a rather long time(hours) rather than as a concentrated jet. This case was studied recently by using the dedicated computercode DECOSIM (DEbris COolability SIMulator) developed at KTH. The code implements the multifluidmodel for liquid-vapor flows in the heat-releasing debris bed and in the volume of pool, where effects ofturbulence are taken into account by the k-epsilon turbulence model. The model used in DECOSIMdescribes the following phenomena: i) filtration of water and vapor in the porous debris bed with heatrelease; ii) turbulent natural convection flows in the water pool; iii) sedimentation of melt particles andtheir interaction with circulatory flow in the pool due to drag and turbulent dispersion; iv) fallout ofparticles, their packing and growth of debris bed. To address the multiscale nature of the problem, acomputationally efficient “gap-tooth” algorithm was developed to speed up considerably the simulationsof long transients typical of gradual melt release mode.Previous DECOSIM simulations of debris bed growth from falling melt particles were performedassuming saturated conditions in the pool at the system pressure. It was shown that the large scale naturalcirculation flows developing in the pool due to vapor production in the porous debris bed, affectsignificantly the debris bed shape because they capture the particles and cause their spreading over thebase-mat of the containment, making the debris bed more flattened and increasing its potentialcoolability. However, an important question remained on how these results would be affected if the poolwater is subcooled. In this case, condensation of vapor in the pool reduces the mixture buoyancy, whichcan decrease the effectiveness of particle spreading.In this work, numerical simulations by DECOSIM code are presented, in which the effects of poolsubcooling on the formation of debris bed are studied. It is shown that, in the subcooled case, thereduction of void due to vapor condensation is offset by the localized heating of water above the debrisbed. The density non-uniformities in the water pool caused by water heating are shown to be sufficient forthe development of large scale natural circulation flows, strong enough for the particle spreading to beefficient. The effect of water subcooling on the ultimate shape of debris bed is demonstrated bycomparing the results obtained for high and low subcooling. This work is performed within the MSWI project supported by APRI/ENSI/NKS

  • 194. Yakush, Sergey
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    In-vessel Debris Bed Coolability and its Influence on the Vessel Failure2013In: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Conference paper (Refereed)
  • 195.
    Yakush, Sergey
    et al.
    A. Yu. Ishlinskii Institute for Problems in Mechanics of the Russian Academy of Sciences.
    Lubchenko, Nazar
    Department of Nuclear Science and Engineering, Massachusetts Institute of Technology.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Risk And Uncertainty Quantification In Debris Bed Coolability2013In: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Pisa, Italy, 2013Conference paper (Refereed)
    Abstract [en]

    An approach to the quantification of uncertainties inherent in the debris bed coolability problem and associated risks analysis is presented. The approach relies on the “Load vs Capacity” concept which compares the thermal loads due to decay heat release and the maximum heat which can be safely removed by water evaporation, both quantities considered as uncertain values due to the uncertainties in the porous material properties, accident scenarios, debris bed shape etc. At the current level of knowledge, not only the distribution functions, but also the ranges of the input parameters cannot be fully quantified. To study the influence of these factors, distribution functions for the Load (decay heat power) and Capacity (maximum power which can be safety removed) are obtained using different ranges and distribution functions for the input parameters. In each case, the probabilities of dryout are evaluated for flat and cone-shaped debris beds. The results obtained demonstrate the advantages of risk-oriented approach and importance of its embedding into the technical decision-making process.

  • 196.
    Yakush, Sergey
    et al.
    A. Yu. Ishlinskii Institute for Problems in Mechanics of the Russian Academy of Sciences.
    Lubchenko, Nazar
    Department of Nuclear Science and Engineering, Massachusetts Institute of Technology.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Surrogate Models for Debris Bed Dryout2013In: The 15th International Topical Meeting on Nuclear Reactor Thermal - Hydraulics, NURETH-15, 2013Conference paper (Refereed)
    Abstract [en]

    The problem of debris bed coolability is important for nuclear power plant severe accident management strategies which employ corium melt fragmentation in a deep pool of water as a means to terminate the accident progression. This work is concerned with development of computationally efficient methods for analysis of corium debris bed coolability. To evaluate the likelihood of severe accident progression due to reheating and remelting of the debris it is important to determine conditions for onset of the dryout in the bed. In the case of a flat one-dimensional debris bed, such conditions are called Dryout Heat Flux (DHF). The DHF is determined by the properties of the bed such as mean particle diameter, porosity, etc., by decay heat power and system pressure and can be obtained analytically. For non-flat configurations of the bed, numerical solution of the multidimensional problem of heat and mass transfer in a porous heat generating media is required in order to find the coolability boundary. In this paper we develop the general functional form of coolability conditions for arbitrary-shape debris beds. It is shown that the DHF concept for a flat debris bed can be extended naturally to multidimensional cases. On the basis of this analysis, a surrogate model is proposed which provides approximation of the coolability boundary, enabling fast calculation of the dryout conditions. The surrogate models enables application of the sensitivity, uncertainty and risk analysis of debris bed coolability.

  • 197. Yericheva, V.A.
    et al.
    Kudinov, Pavel
    Semianalytic method for investigation of aeroelastic resonant oscillations of profile in flow of viscid gas2005In: Proceedings of XII International Symposium “Methods of Discrete Singularities in Problems of Mathematical Physics", 2005, p. 126-130Conference paper (Refereed)
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