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  • 1. Buck, M.
    et al.
    Buerger, M.
    Miassoedov, A.
    Gaus-Liu, X.
    Palagin, A.
    Godin-Jacqmin, L.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Chudanov, V.
    The LIVE program: Results of test L1 and joint analyses on transient molten pool thermal hydraulics2010Inngår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, nr 1, s. 46-60Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e.g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO3-NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE activities also provide data for a better understanding of in-core corium pool behaviour. The experimental results are being used for the development and validation of mechanistic models for the description of molten pool behaviour, In the present paper, a range of different models is used for post-test calculations and comparative analyses. This includes simplified, but fast running models implemented in the severe accident codes ASTEC and ATHLET-CD. Further, a computational tool developed at KTH (PECM model implemented in Fluent) is applied. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE (Nuclear Safety Institute of Russian Academy) within the RASPLAV project and was further improved within the ISTC 2936 Project.

  • 2.
    Cadinu, Francesco
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Analysis of In-Vessel Coolability and Retention with Control Rod Guide Tube Cooling in Boiling Water Reactors2009Konferansepaper (Fagfellevurdert)
  • 3.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The Effect of Internal Pressure and Debris Bed Thermal Properties on BWR Vessel Lower Head Failure and Timing2013Inngår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 4.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Experimental Investigation of Melt Debris Agglomeration with High Melting Temperature Simulant Materials2009Inngår i: Proceedings of implementation of severe accident management measures (ISAMM-2009), Paul Scherrer Institut , 2009Konferansepaper (Fagfellevurdert)
  • 5.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Agglomeration and size distribution of debris in DEFOR-A experiments with Bi2O3-WO3 corium simulant melt2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 263, s. 284-295Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Flooding of lower drywell has been adopted as a cornerstone of severe accident management strategy in Nordic type Boiling Water Reactors (BWR). It is assumed that the melt ejected into a deep pool of water will fragment, quench and form a porous debris bed coolable by natural circulation. If debris bed is not coolable, then dryout and possibly re-melting of the debris can occur. Melt attack on the containment basemat can threaten containment integrity. Agglomeration of melt debris and formation of solid "cake" regions provide a negative impact on coolability of the porous debris bed. In this work we present results of experimental investigation on the fraction of agglomerated debris obtained in the process of hot binary oxidic melt pouring into a pool of water. The Debris Bed Formation and Agglomeration (DEFOR-A) experiments provide data about the effects of the pool depth and water subcooling, melt jet diameter, and initial melt superheat on the fraction of agglomerated debris. The data presents first systematic study of the debris agglomeration phenomena and facilitates understanding of underlying physics which is necessary for development and validation of computational codes to enable prediction of the debris bed coolability in different scenarios of melt release.

  • 6.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The DEFOR-A Experiment on Fraction of Agglomerated Debris as a Function of Water Pool Depth2010Inngår i: The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), 2010Konferansepaper (Fagfellevurdert)
  • 7.
    Torregrosa, Claudio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Coupled 3D Thermo-mechanical Analysis of a Nordic BWR Vessel Failure and Timing2013Inngår i: Proceedings 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 8.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Critical Heat Flux CorrelationsManuskript (preprint) (Annet vitenskapelig)
  • 9.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Development, validation and application of an effective convectivity model for simulation of melt pool heat transfer in a light water reactor lower head2007Licentiatavhandling, monografi (Annet vitenskapelig)
    Abstract [en]

    Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. The research is also focusing on evaluation of the proposed accident management schemes for mitigating the consequences of such accidents.

    During a hypothetical severe accident, whatever the scenario, there is likelihood that the core material will be relocated and accumulated in the lower plenum in the form of a debris bed or a melt pool. Physical phenomena involved in a severe accident progression are complex. The interactions of core debris or melt with the reactor structures depend very much on the debris bed or melt pool thermal hydraulics. That is why predictions of heat transfer during melt pool formation in the reactor lower head are important for the safety assessment.

    The main purpose of the present study is to advance a method for describing turbulent natural convection heat transfer of a melt pool, and to develop a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of Core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident.

    Given the insights gained from Computational Fluid Dynamics (CFD) simulations, a physics-based model and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a Boiling Water Reactor (BWR) during the late phase of an in-vessel core melt progression. A model is developed for the core debris bed heat up and formation of a melt pool in the lower head of the reactor vessel, and implemented in a commercial CFD code. To describe the natural convection heat transfer inside the volumetrically decay-heated melt pool, we advanced the Effective Convectivity Conductivity Model (ECCM), which was previously developed and implemented in the MVITA code. In the present study, natural convection heat transfer is accounted for by only the Effective Convectivity Model (ECM). The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables simulations of heat transfer of a high Rayleigh melt pool in 3D large dimension geometry.

    In order to describe the phase-change heat transfer associated with core debris, a temperature-based enthalpy formulation is employed in the ECM (the phase-change ECM or so called the PECM). The PECM is capable to represent possible convection heat transfer in a mushy zone. The simple approach of the PECM method allows implementing different models of the fluid velocity in a mushy zone for a non-eutectic mixture. The developed models are validated by a dual approach, i.e., against the existing experimental data and the CFD simulation results.

    The ECM and PECM methods are applied to predict thermal loads to the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heat up and melting in the BWR lower plenum. Applying the ECM and PECM to simulations of reactor-scale melt pool heat transfer, the results of the ECM and PECM calculations show an apparent effectiveness of the developed methods that enables simulations of long term accident transients. It is also found that during severe accident progression, the cooling by water flowing inside the CRGTs plays a very important role in reducing the thermal load on the reactor vessel wall. The results of the CFD, ECM and PECM simulations suggest a potential of the CRGT cooling as an effective mitigative measure during a severe accident progression.

  • 10.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head2009Doktoravhandling, med artikler (Annet vitenskapelig)
    Abstract [en]

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents.  In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment.  The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis.

     

    The CFD method, on the one hand, is indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling plays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs.

  • 11.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    An effective convectivity model for simulation of in-vessel core melt progression in a boiling water reactor2008Inngår i: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work", 2008, s. 925-935Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The present paper is concerned with development and application of a so-called Effective Convectivity Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to simulate heat transfer of an internally heated volume in 3D complex geometry is demonstrated through examples of heat transfer analysis in a BWR lower plenum being cooled by coolant flow in Control Rod Guide Tubes. Simulation results and key findings of this case are reported and discussed.

  • 12.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Analysis of melt pool heat transfer in a BWR lower head2006Inngår i: Transactions of the American Nuclear Society, 2006, s. 629-631Konferansepaper (Fagfellevurdert)
  • 13.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Application of the phase-change effective convectivity model to analysis of core melt pool formation and heat transfer in a BWR lower head2008Inngår i: Trans Am Nucl Soc, 2008, s. 617-618Konferansepaper (Fagfellevurdert)
  • 14.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A Synergistic use of CFD, Experiments and Effective Convectivity Model to Reduce Uncertainty in BWR Severe Accident Analysis2010Konferansepaper (Fagfellevurdert)
  • 15.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Local Heat Transfer From The Corium Melt Pool to the BWR Vessel Wall2011Inngår i: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
  • 16.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head2009Inngår i: International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009, Atomic Energy Society of Japan , 2009, Vol. 2, s. 1523-1537Konferansepaper (Fagfellevurdert)
    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.

  • 17.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effective convectivity model for simulation of molten metal layer heat transfer in a boiling water reactor lower head2013Inngår i: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, Vol. 2013, s. 231501-Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 18.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head2010Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, nr 9, s. 2148-2159Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 19.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An approach to numerical simulation and analysis of molten corium coolability in a BWR lower head2008Konferansepaper (Fagfellevurdert)
  • 20.
    Tran, Chi Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    A Study on the Integral Effect of Corium Material Properties on Melt Pool Heat Transfer in a Boiling Water Reactor2011Inngår i: Proceedings 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
  • 21.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Simulation of core melt pool formation in a reactor pressure vessel lower head using an effective convectivity model2007Inngår i: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12, 2007Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Boiling Water Reactor (BWR) Lower Plenum (LP) during a hypothetical severe accident scenario. The Phase-change ECM (PECM) was examined using a dual-tier approach, namely validations against existing experimental data and validation against results obtained from Computational Fluid Dynamics (CFD) simulations. Three models of the fluid velocity in a mushy zone were implemented in the PECM and their performance was compared. A linear dependency of the fluid velocity in a mushy zone on the fluid fraction applied in the PECM gives consistent results with the CFD simulations. The PECM was applied to simulation of heat transfer in a BWR lower plenum during a severe accident scenario. Results of the PECM core melt pool transient simulation are reported in the paper. We show that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

  • 22.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Simulation of core melt pool formation in a reactor pressure vessel lower head using an effective convectivity model2009Inngår i: Nuclear engineering and technology : an international journal of the Korean Nuclear Society, ISSN 1738-5733, E-ISSN 2234-358X, Vol. 41, nr 7, s. 929-944Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a Mushy Zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity On fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The Study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

  • 23.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation2009Inngår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, nr 8, s. 849-859Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper, and its companion paper [Tran C.T., Dinh, IN. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application. Progress in Nuclear Energy (companion paper), in preparation] document the development, validation and applications of a simulation platform for computationally-effective, sufficiently-accurate numerical predictions of core melt-structure-water interactions in the light water reactor lower head during a postulated severe core-melting accident. The centerpiece of this work is the Effective Convectivity Model (ECM) for description of energy splitting in a core melt pool. Built on the concept of characteristic velocities in Effective Convectivity Conductivity Model and supported by the key findings obtained from Computational Fluid Dynamics (CFD) simulations of turbulent natural convection, heat transfer and phase changes in volumetrically heated liquid pools, the ECM is refined and extended to three-dimensions and phase changes to enable simulations of melt pool formation and corium coolability in complex geometry such as a Boiling Water Reactor (BWR) lower plenum.

  • 24.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application2009Inngår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, nr 8, s. 860-871Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

  • 25.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A Computational Study On Instrumentation Guide Tube Failure During a Severe Accident in Boiling Water Reactor2011Inngår i: The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
  • 26.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Analysis of Instrumentation Guide Tube Failure in a BWR Lower Head2012Inngår i: Proc. 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Konferansepaper (Fagfellevurdert)
  • 27.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Assessment with Coupled Thermo-Mechanical Creep Analysis of Combined CRGT and External Vessel Cooling Efficiency for a BWR2011Inngår i: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
  • 28.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head2012Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 249, s. 146-153Artikkel i tidsskrift (Fagfellevurdert)
    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.

1 - 28 of 28
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