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  • 1. Bandini, G.
    et al.
    Polidori, M.
    Gerschenfeld, A.
    Pialla, D.
    Li, S.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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.
    Huber, K.
    Cheng, X.
    Bruzzese, C.
    Class, A. G.
    Prill, D. P.
    Papukchiev, A.
    Geffray, C.
    Macian-Juan, R.
    Maas, L.
    Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors2015In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 281, p. 22-38Article in journal (Refereed)
    Abstract [en]

    The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

  • 2. Buck, M.
    et al.
    Buerger, M.
    Miassoedov, A.
    Gaus-Liu, X.
    Palagin, A.
    Godin-Jacqmin, L.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Chudanov, V.
    The LIVE program: Results of test L1 and joint analyses on transient molten pool thermal hydraulics2010In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, no 1, p. 46-60Article in journal (Refereed)
    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.

  • 3. Buerger, M.
    et al.
    Buck, M.
    Pohlner, G.
    Rahman, S.
    Kulenovic, R.
    Fichot, F.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Miettinen, J.
    Lindholm, I.
    Atkhen, K.
    Coolability of particulate beds in severe accidents: Status and remaining uncertainties2010In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, no 1, p. 61-75Article in journal (Refereed)
    Abstract [en]

    Particulate debris beds may form during different stages of a severe core melt accident; e.g. in the degrading hot core, due to thermal stresses during reflooding, in the lower plenum, by melt flow from the core into water in the lower head, and in the cavity by melt flow out of a failing RPV into a wet cavity. Deep water pools in the cavity are used in Nordic BWRs as an accident management measure aiming at particulate debris formation and coolability. It has been elaborated in the joint work of the European Severe Accident Research Network (SARNET) in Work Package (WP) 11.1 that coolability of particulate debris, reflooding of hot debris as well as boil-off under decay heat (long-term coolability), is strongly favoured by 2D/3D effects in beds with non-homogeneous structure and shape. Especially, water inflow from the sides and via bottom regions strongly improves coolability as compared to 1D situations with top flooding, the latter being in the past the basis of analyses on coolability. Data from experiments included in the SARNET network (DEBRIS at IKE and STYX at VTT) and earlier ones (e.g. POMECO at KTH) have been used to validate key constitutive laws in 2D codes as WABE (IKE) and ICARE/CATHARE (IRSN), especially concerning flow friction and heat transfer. Major questions concern the need of the explicit use of interfacial friction to adequately treat the various flow situations in a unified approach, as well as the adequate characterization of realistic debris composed of irregularly shaped particles of different sizes. joint work has been supported by transfer of the WABE code to KTH and VTT. Concerning realistic debris, the formation from breakup of melt jets in water is investigated in the DEFOR experiments at KTH. Present results indicate that porosities in the debris might be much higher than previously assumed, which would strongly support attainment of coolability. Calculations have been performed with IKEJET/IKEMIX describing jet breakup, mixing and settling of resulting particles. Models about debris bed formation and porosity are developed at KTH. The codes have been applied to reactor conditions for analysing the potential for coolability in the different phases of a severe accident. Calculations have been performed with WABE (MEWA) implemented in ATHLET-CD and with ICARE/ICATHARE for degraded cores and debris beds in the lower plenum, under reflooding and boil-off. Ex-vessel situations have also been analysed. Strong effects of lateral water inflow and cooling by steam in hot areas have been demonstrated. In support, some typical basic configurations have been analysed, e.g. configurations with downcomers considered as possible AM measures. Melt pool formation or coolability of particulate debris is a major issue concerning melt retention in the core and the lower head. Present conclusions from those analyses for adequate modelling in ASTEC are outlined as well as remaining uncertainties. Experimental and analysis efforts and respective continued joint actions are discussed, which are needed to reach resolution of the coolability issue.

  • 4.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

    In a previous study [1], the US NRC code TRACE was employed to simulate the FIX-II tests which were carried out to investigate the loss of coolant accident (LOCA) of a boiling water reactor (BWR). Results exhibited that the TRACE simulation was sensitive to modelling parameters. In order to further qualify the TRACE code for BWR safety analysis, and to increase our confidence in the simulation results, sensitivity and uncertainty analysis is performed in this paper for the possible uncertain parameters, so as to identify the most influential ones. 12 parameters related to the simulated physical phenomena are selected by resorting to phenomena identification and ranking tables (PIRTs) in relative references. The sensitivity analysis method chosen is based on Finite Mixture Models (FMM) together with Hellinger distance and Kullback-Leibler divergence. Kolmogorov-Smirnov test is first introduced to combine FMM, and it has better performance in screening. Sensitivity analysis results of FMM method show that decay power, choked flow multipliers and break area have the most important influence on calculating peak cladding temperature (PCT). Although previous study failed to predict PCT, uncertainty analysis provides a certain range that successfully covers experiment result.

  • 5.
    Chen, Yaodong
    et al.
    State Nuclear Power Research Institute, United States .
    Weimin, Ma
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Cui, L.
    Numerical investigation of Fukushima Daiichi-2 SBO scenario2014In: International Congress on Advances in Nuclear Power Plants, ICAPP 2014, 2014, Vol. 2, p. 995-1003Conference paper (Refereed)
    Abstract [en]

    Simulations of the severe accident progression for Fukushima Daiichi NPP Unit 2 (1F2) are performed using the MELCOR code. Detailed modeling of the plant is developed to represent the whole reactor system and its safety systems. The predicted results are compared with the plant data measured during the accident. By applying the main actions taken during the accident and the assumptions into the full plant MELCOR modeling, the major physical phenomena from core uncovery and degradation till reflooding of reactor core by fire pump injection are reproduced in the simulations. The trend of simulation results agree in general with the limited data (e.g., pressures) measured by the plant. The closed RCIC cycle, which involved steam flow and working process, and its interacting with reactor cooling status was modeled by user defined control function in the simulation. The simulations reveal that: The operations of RCIC kept the reactor core flooded to the top for more than 70 hours after the earthquake until the suppression pool water got saturated. Sea water might have flooded into the TORUS room to more extent than as assumed, which kept cooling of suppression pool, and delayed the failure of RCIC. Around 2 hours before the cooling water by fire pump was able to inject water into the reactor, the core damage started at around 76.5hr and got oxidized severely within 2 hours. While no further degradation occurred, the core geometry was maintained, and capable of being cooled by sea water injection.. A leakage has possibly occurred somewhere in RCS steam phase region, to account for pressurization of containment dry well before suppression pool got saturated.

  • 6. Cheng, X.
    et al.
    Batta, A.
    Bandini, G.
    Roelofs, F.
    Van Tichelen, K.
    Gerschenfeld, A.
    Prasser, M.
    Papukchiev, A.
    Hampel, U.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems2015In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, p. 2-12Article in journal (Refereed)
    Abstract [en]

    Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

  • 7. Chikhi, N.
    et al.
    Coindreau, O.
    Li, L. X.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Taivassalo, V.
    Takasuo, E.
    Leininger, S.
    Kulenovic, R.
    Laurien, E.
    Evaluation of an effective diameter to study quenching and dry-out of complex debris bed2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 24-41Article in journal (Refereed)
    Abstract [en]

    Many of the current research works performed in the SARNET-2 WP5 deal with the study of the coolability of debris beds in case of severe nuclear power plant accidents. One of the difficulties for modeling and transposition of experimental results to the real scale and geometry of a debris bed in a reactor is the difficulty to perform experiments with debris beds that are representative for reactor situations. Therefore, many experimental programs have been performed using beds made of multi-diameter spheres or non-spherical particles to study the physical phenomena involved in debris bed coolability and to evaluate an effective diameter. This paper first establishes the ranges of porosity and particle size distribution that might be expected for in-core debris beds and ex-vessel debris beds. Then, the results of pressure drop and dry-out heat flux (DHF) measurements obtained in various experimental setups, POMECO, DEBRIS, COOLOCE/STYX and CALIDE/PRELUDE, are presented. The issues of particle size distribution and non-sphericity are also investigated. It is shown that the experimental data obtained in "simple" debris beds are relevant to describe the behavior of more complex beds. Indeed, for several configurations, it is possible to define an "effective" diameter suitable for evaluating (with the porosity) some model parameters as well as correlations for the pressure drop across the bed, the steam flow rate during quenching and the DHF.

  • 8.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, T.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Space-time convergence study based on OECD Ringhals-1 Stability Benchmark2012In: Transactions of the American Nuclear Society: Volume 106, 2012, 2012, p. 984-987Conference paper (Refereed)
  • 9.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    University of Illinois, United States .
    Sensitivity analysis of input uncertain parameters on BWR stability using TRACE/PARCS2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 67, p. 49-58Article in journal (Refereed)
    Abstract [en]

    The unstable behavior of Boiling Water Reactors (BWR), which is known to occur at certain power and flow conditions, could cause SCRAM and decrease the economic performance of the plant. For better prediction of BWR stability and understanding of influential parameters, two TRACE/PARCS models of Ringh-als-1 and Oskarshamn-2 BWRs were employed to perform a sensitivity study. Using the propagation of input errors uncertainty method's results, an attempt has been made to identify the most influential parameters affecting the stability. Furthermore, a methodology using the spearman rank correlation coefficient has been used to identify the most influential parameters on the stability parameters (decay ratio and frequency).

  • 10.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    Univ Illinois, USA.
    Space–time convergence analysis on BWR stability using TRACE/PARCS2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 51, p. 295-306Article in journal (Refereed)
    Abstract [en]

    Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  • 11. Gong, S.
    et al.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Wang, C.
    Mei, Y.
    Gu, H.
    An investigation on dynamic thickness of a boiling liquid film2015In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 90, p. 636-644Article in journal (Refereed)
    Abstract [en]

    Motivated by understanding the micro-hydrodynamics of boiling heat transfer and the mechanism of critical heat flux (CHF) occurrence, the present study is to investigate the boiling phenomenon in a liquid film whose dynamic thickness is recorded by a confocal optical sensor with the measurement accuracy of micrometres, while the bubble dynamics of the boiling in the film is visualized by a high-speed photography. This paper is focused on a statistical analysis of the measured thickness signals for the boiling condition ranging from low heat flux to high heat flux (near or at CHF). The dynamic thickness of liquid film appears oscillating with peak values, resulting from the liquid film movements due to nucleation of bubble(s) and its growth and rupture. The statistical analysis in a certain period indicates there emerge three distinct liquid film thickness ranges: 0-50 μm, 50-500 μm and 500-2500 μm, seemingly corresponding to the microlayer, macrolayer and bulk layer. With increasing heat flux to a specific extent, the bulk layer disappears, and then the macrolayer gradually decreases to ∼105 μm, beyond which the liquid film may lose its integrity and CHF occurs at 1.563 MW/m2.

  • 12.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Liangxing
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Experimental Study on Bubble Dynamics of High-Heat-Flux Boiling in a Horizontal Liquid Layer2011In: Proceedings of the 2011 international congress on advances in nuclear power plants: ICAPP2011, American Nuclear Society, 2011Conference paper (Refereed)
  • 13.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Experimental Study on Boiling Phenomenon in a Liquid Layer2010In: 7th International Conference on Multiphase Flow - ICMF 2010 Proceedings, International Conference on Multiphase Flow (ICMF) , 2010Conference paper (Refereed)
    Abstract [en]

    This work investigates boiling phenomena by means of imaging and characterization of bubble dynamics in the vicinity of the bubble’s nucleation site. A silicon wafer is used as heat transfer surface so that MEMS fabrication can be applied to create artificial cavity for prescribed nucleation site. The well-controlled bubbles growing on such nucleation site can facilitate measurement and observation. High-speed video camera is employed in visualization, and the instantaneous thickness of the liquid layer is recorded by a confocal optical sensor. Tests are first performed on a water layer with the thickness of 7.5mm±0.5mm, and the bubble departure diameter and frequency as well as the transient evolution of bubble diameter and foot size are obtained in isolated bubble regime. Bubble departure diameter enlarges with increasing heat flux, and the measured maximum diameter is around 3.2 mm. With the decrease of the liquid layer thickness to 2 mm, the bubbles are found to remain on the heater surface for a relative long period, with a dry spot initiation under the bubble that becomes rewetted after the bubble bursting. As the water layer thickness decreases further, irreversible dry spot appears, suggesting a minimum “safe” film thickness in the range from 1.2 to 1.9 mm under the tested heat flux range from 26 kW/m2 to 52 kW/m2.

  • 14.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study of rupture dynamics of evaporating liquid films on different heater surfaces2011In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 54, no 7-8, p. 1538-1547Article in journal (Refereed)
    Abstract [en]

    Experimental data were obtained to reveal the complex dynamics of thin liquid films evaporating on heated horizontal surfaces, including formation and expansion of dry spots that occur after the liquid films decreased below critical thicknesses. The critical thickness of water film evaporating on various material surfaces is measured in the range of 60-150 mu m, increasing with contact angle and heat flux while decreasing with thermal conductivity of the heater material. In the case of hexane evaporating on a titanium surface, the liquid film is found resilient to rupture, but starts oscillating as the averaged film thickness decreases below 15 mu m.

  • 15.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Diagnostic techniques for the dynamics of a thin liquid film under forced flow and evaporating conditions2010In: MICROFLUID NANOFLUID, ISSN 1613-4982, Vol. 9, no 6, p. 1077-1089Article in journal (Refereed)
    Abstract [en]

    Motivated by quantification of micro-hydrodynamics of a thin liquid film which is present in industrial processes, such as spray cooling, heating (e.g., boiling with the macrolayer and the microlayer), coating, cleaning, and lubrication, we use micro-conductive probes and confocal optical sensors to measure the thickness and dynamic characteristics of a liquid film on a silicon wafer surface with or without heating. The simultaneous measurement on the same liquid film shows that the two techniques are in a good agreement with respect to accuracy, but the optical sensors have a much higher acquisition rate up to 30 kHz which is more suitable for rapid process. The optical sensors are therefore used to measure the instantaneous film thickness in an isothermal flow over a silicon wafer, obtaining the film thickness profile and the interfacial wave. The dynamic thickness of an evaporating film on a horizontal silicon wafer surface is also recorded by the optical sensor for the first time. The results indicate that the critical thickness initiating film instability on the silicon wafer is around 84 mu m at heat flux of similar to 56 kW/m(2). In general, the tests performed show that the confocal optical sensor is capable of measuring liquid film dynamics at various conditions, while the micro-conductive probe can be used to calibrate the optical sensor by simultaneous measurement of a film under quasi-steady state. The micro-experimental methods provide the solid platform for further investigation of the liquid film dynamics affected by physicochemical properties of the liquid and surfaces as well as thermal-hydraulic conditions.

  • 16.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Measurement of Film Dynamics in a Boiling Liquid Film2011In: Proceedings of NURETH-14 The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics, 2011Conference paper (Refereed)
  • 17.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Measurement of thin liquid film dynamics under forced flow and evaporating condition2009In: Proceeding of ECI International Conference on Boiling Heat Transfer, 2009Conference paper (Refereed)
  • 18.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics.
    Simulation and validation of the dynamics of liquid films evaporating on horizontal heater surfaces2012In: Applied Thermal Engineering, ISSN 1359-4311, E-ISSN 1873-5606, Vol. 48, p. 486-494Article in journal (Refereed)
    Abstract [en]

    In this study a non-linear governing equation based on lubrication theory is employed to model the thinning process of an evaporating liquid film and ultimately predict the critical thickness of the film rupture under impacts of various forces resulting from mass loss, surface tension, gravity, vapor recoil and thermo-capillary. It is found that the thinning process in the experiment is well reproduced by the simulation. The film rupture is caught by the simulation as well, but it underestimates the measured critical thickness at the film rupture. The reason may be that the water wettability of the heater surfaces is not taken into account in the model. Thus, the minimum free energy criterion is used to obtain a correlation which combines the contact angle (reflection of wettability) with the critical thickness from the simulation. The critical thicknesses predicted by the correlation have a good agreement with the experimental data (the maximum deviation is less than 10%).

  • 19. Gong, Shengjie
    et al.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gu, Hanyang
    An experimental investigation on bubble dynamics and boiling crisis in liquid films2014In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 79, p. 694-703Article in journal (Refereed)
    Abstract [en]

    This paper presents an experimental study of boiling and boiling crisis in a liquid film on a heater surface. The critical heat flux (CHF) values obtained in the present experiment mirror that of pool boiling, irrespective of initial liquid film thickness and liquid supply rate in the liquid film boiling case. This observation reinforces to the "scale separation" concept that high-heat-flux boiling and burnout are governed by micro-hydrodynamics in the liquid film on the heater surface. In addition to the CHF data, evolutions of bubbles and dry spots in the boiling liquid film are captured by means of high-speed high-resolution video camera. The dry spots were observed over surface heat flux ranging from 0.3 MW/m(2) to CHF, typically covering an area less than 10% of the heater surface. Three types of dry spot evolution are observed: (1) under the low heat flux, dry spots are rewetted by receding water dam upon rupture of corresponding bubbles; (2) as the heat flux reaches 1.25 MW/m(2), dry spots rewetting is additionally aided by liquid flow driven by growth of bubbles nucleated in the vicinity; (3) upon approaching the CHF, dry spot(s) cannot be rewetted anymore and expand laterally, leading to boiling crisis (burnout of the heater surface). The richness of observations and characterization of micro-hydrodynamics in the present study further demonstrates that observations and measurements on boiling liquid films provide a paramount window for investigation and understanding of physical mechanisms of boiling and boiling crisis.

  • 20.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Liangxing
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on boiling phenomena in a liquid layerIn: International journal of thermal sciences, ISSN 1290-0729, E-ISSN 1778-4166Article in journal (Other academic)
  • 21.
    Gong, Shengjie
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Liangxing
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on the effect of liquid film thickness on bubble dynamics2013In: Applied Thermal Engineering, ISSN 1359-4311, E-ISSN 1873-5606, Vol. 51, no 1-2, p. 459-467Article in journal (Refereed)
    Abstract [en]

    Experiments were conducted to investigate the boiling phenomenon in various liquid layers on a silicon heater surface with an artificial cavity. Deionized water is employed as working liquid. The emphasis is placed on how the liquid layer thickness affects bubble behaviour and liquid layer integrity for nucleate boiling under the isolated bubble regime. The experimental results show that for boiling in a liquid layer of ∼7.5 mm, the bubble dynamics reproduce the typical pool boiling characteristics with the averaged maximum diameter of 3.2 mm for the isolated bubbles growing on the cavity. As the water layer thickness decreases to the level comparable with the bubble departure diameter, the bubble is found to remain on the heater surface for an extended period, with a dry spot forming under the bubble but rewetted after the bubble rupture occurs. Further reducing the liquid layer thickness, an irreversible dry spot appears, suggesting a minimum rewettable thickness ranging from 1.2 mm to 1.9 mm corresponding to heat flux of 26 kW/m2 to 52 kW/m2. The void measured in the cavity confirms that it is dry inside the artificial cavity at high heat flux.

  • 22. Gu, H.
    et al.
    Wang, C.
    Gong, S.
    Mei, Y.
    Li, H.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Investigation on contact angle measurement methods and wettability transition of porous surfaces2016In: Surface & Coatings Technology, ISSN 0257-8972, E-ISSN 1879-3347, Vol. 292, p. 72-77Article in journal (Refereed)
    Abstract [en]

    Various solid surfaces (e.g., smooth titanium surface, smooth aluminum surface, polished copper surfaces, polished silver surfaces and porous copper surfaces) were prepared to quantify the reliability of half-angle algorithm and axisymmetric drop shape analysis (ADSA) algorithm for calculating contact angles. Besides, the effects of surface conditions on contact angle values were also investigated. The experimental results of 10 repeated tests for each surface show that both algorithms have good accuracy for an acute contact angle, while the ADSA algorithm is better than the half-angle algorithm for an obtuse contact angle. Furthermore, with the decrease of surface roughness, the contact angle increases but the standard deviation of contact angles by 10 repeated tests decreases. In addition, the porous layer on copper surface by electrochemical deposition shows a super hydrophilic property, but it could change to be super hydrophobic after exposed in ambient air for 24 h. Interestingly, the surface wettability reverses to be super hydrophilic again after it is immersed in water, and the inorganic contamination is the reason of formal change from the super hydrophilic status to the super hydrophobic status.

  • 23. Hu, X.
    et al.
    Xing, M.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Trace validation against loca transients performed on fix-II facility2013In: 2013 21st International Conference on Nuclear Engineering: Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled CodesChengdu, China, July 29–August 2, 2013, ASME Press, 2013Conference paper (Refereed)
    Abstract [en]

    As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Lossof- coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermalhydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.

  • 24.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 83-92Article in journal (Refereed)
    Abstract [en]

    Motivated to investigate the thermal hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3 days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13 m and 15 m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

  • 25.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation and application of the MEWA code to analysis of debris bed coolability2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, p. 22-37Article in journal (Refereed)
    Abstract [en]

    This paper was first aimed at validating the MEWA code against experiments for two-phase flow and dryout in particulate beds, and then investigating the coolability of ex-vessel debris beds with cylindrical, conical and truncated conical shapes assumed to form under severe accident scenarios of a boiling water reactor. The validation was mainly performed against the POMECO-FL and POMECO-HT experiments carried out at KTH for investigating frictional laws and coolability limit (dryout) of particulate beds, respectively. The comparison of the experimental and numerical results shows that the MEWA code is capable of predicting both the pressure drop of two-phase flow through porous media and the dryout condition of various stratified beds. While the coolability of a one-dimensional homogeneous debris bed is bounded by counter-current flow limit (CCFL), the coolability of a heap-like debris bed can be improved due to lateral ingression of coolant in a multi-dimensional geometry. The simulations showed that the dryout power density of a prototypical debris bed was roughly inversely proportional to the bed's height regardless of the bed's shape. The impacts of a debris bed's features on coolability are manifested in three aspects: multidimensionality and contour surface area of the bed, as well as the uniformity of its shape. The contour surface area is defined as the interface between debris bed and water pool, and its effective value depends on the surface orientation that determines the amount of water ingress and vapor escape. The perfect uniformity in bed's shape as cylindrical bed results in even distributions of temperature and void fraction. The dryout power density was also predicted to be strongly correlated to the uniformity of bed's shape. The MEWA simulation also predicted that coolability was improved by an downcomer embedded in the center of debris bed. The efficiency of such enhancement was largely determined by the downcomer's length, whose optimal value was obtained in simulation.

  • 26.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of the MEWA code agsinst POMECO-HT experiments and cool ability analysis of stratified debris BEDS2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, 2015, Vol. 4, p. 3279-3291Conference paper (Refereed)
    Abstract [en]

    Motivated by qualification of the MEWA code for coolability analysis of debris beds formed during severe accidents of light water reactors, the present work presents a validation of the code against the experimental data obtained on the POMECO-HT facility for investigation of two-phase flow and heat transfer limits in particulate beds with various characteristics. The volumetrically heated particulate beds used in the POMECO-HT experiment are packed in various configurations, including homogeneous bed, radially stratification, triangular stratification, axial stratification, and multi-stratification. To investigate coolability enhancement by bottom-fed induced natural circulation, a downcomer is employed. Besides, the influence of the interfacial drag is also studied. The results show that simulation results of the MEWA code is overall comparable with the experimental data in term of dryout conditions of the particulate beds. For the 1-D top-flood case, the dryout heat flux is mainly determined by counter-current flow limit. While for certain cases the multidimensionality may help to break CCFL. Besides, the debris bed’s coolabiltiy can be significantly improved due to the natural circulation flow from the bottom induced by using downcomer. The interfacial drag affects the coolability by means of varying the pressure field inside the bed. For the top-flood case, the dryout condition deteriorates since the vapor and coolant flow reversely and thus the interfacial drag increases the flow resistance. Whereas for the bottom-fed case, the dryout heat flux rises remarkably when considering the interfacial drag, because the vapor and coolant flow in the same direction and the interfacial drag helps to pull coolant upward from the bottom.

  • 27. Journeau, C.
    et al.
    Bonnet, J. M.
    Godin-Jacqmin, L.
    Piluso, P.
    Tarabelli, D.
    Dufour, E.
    Spindler, B.
    Nicolas, L.
    Altstadt, E.
    Atkhen, K.
    Dutheillet, Y.
    Lamy, J. S.
    Bandini, G.
    Ederli, S.
    Barrachin, M.
    Cranga, M.
    Duriez, C.
    Fichot, F.
    Repetto, G.
    Koundy, V.
    Birchley, J.
    Bottomley, D.
    Wiss, T.
    Buck, M.
    Burger, M.
    Cheynet, B.
    Dimov, D.
    Grudev, P.
    Stefanova, A.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sehgal, Balraj
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Drath, T.
    Hollands, T.
    Kleinhietpass, I.
    Koch, M.
    Duspiva, J.
    Kiselova, M.
    Kujal, B.
    Vokac, P.
    Erdmann, W.
    Müller, C.
    Spengler, C.
    Fischer, M.
    Schmidt, W.
    Hellmann, S.
    Foit, J. J.
    Miassoedov, A.
    Steinbriick, M.
    Stuckert, J.
    Gallego, E.
    Martin, M. G.
    Meleg, T.
    Ohai, D.
    Matejovic, P.
    Mathew, M.
    Sdouz, G.
    Sevon, T.
    European Research on the Corium issues within the SARNET network of excellence2008In: International Conference on Advances in Nuclear Power Plants, ICAPP 2008, 2008, p. 1172-1181Conference paper (Refereed)
    Abstract [en]

    Within SARNET, the corium topic covers all the behaviors of corium from early phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission product release. The corium topic regroups in three work packages the critical mass of competence required to improve significantly the corium behavior knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better efficiency and to rationalize the R&D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of experiments, while, in the next periods, a larger focus will be given to integral experiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This paper summarizes the progress that has been achieved in the frame of the networking activities. A special focus is placed on the melt pool and debris coolability and corium-concrete interaction, in which, the effects due to multidimensional geometries and heterogeneities has been shown, during SARNET, to play a crucial role and for which further research is still needed.

  • 28. Journeau, C.
    et al.
    Bouyer, V.
    Cassiaut-Louis, N.
    Fouquart, P.
    Piluso, P.
    Ducros, G.
    Gossé, S.
    Guéneau, C.
    Quaini, A.
    Fluhrer, B.
    Miassoedov, A.
    Stuckert, J.
    Steinbrück, M.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sehgal, Bal Raj
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Hozer, Z.
    Guba, A.
    Manara, D.
    Bottomley, D.
    Fischer, M.
    Langrock, G.
    Schmidt, H.
    Kiselova, M.
    Ždarek, J.
    Safest roadmap for corium experimental research in Europe2018In: ASCE-ASME Journal of Risk and Uncertainty in Engineering Systems, Part B: Mechanical Engineering, ISSN 2332-9017, Vol. 4, no 3, article id 030901Article in journal (Refereed)
    Abstract [en]

    Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.

  • 29. Journeau, Christophe
    et al.
    Cranga, Michel
    Foit, Jerzy
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grudev, Pavlin
    A European joint work plan on molten core concrete interaction2009In: International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009, Atomic Energy Society of Japan , 2009, p. 1058-1065Conference paper (Refereed)
    Abstract [en]

    In the case of a severe accident with vessel melt-through, the Molten Core Concrete Interaction (MCCI) may take place and become a threat to the integrity of the containment, which is the ultimate barrier between the corium and the environment. During SARNET project (EURATOM 6th Framework programme) 41/2 years, experimental programs related to MCCI issues have been pursued on the VULCANO, HECLA, COMETA and ARTEMIS facilities and several benchmark exercises have been conducted, in particular on reactor scale applications. This issue has been considered within the SARNET project (2004-2008) to have a high priority. Within the new project SARNET2 (2009-2013), a significant experimental program on MCCI, coupled with joint interpretation, modelling and code applications, has been proposed in view of understanding these phenomena and towards closing the safety issue. This research program has been designed to ensure complementarity with the ongoing MCCI-2 project of the OECD-NEA. It will address the following issues: effect of the concrete nature on 2D ablation profiles, role of the metallic layer on the MCCI, and efficiency of water cooling to terminate the ablation of concrete. It will also have a specific focus on the transposition of R&D results to the reactor scale. More than 17 European organizations will contribute to these activities, which will include a significant experimental program both in simulant and in prototypic materials.

  • 30.
    Karbojian, Aram
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Davydov, Mikhail
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A scoping study of debris formation in DEFOR experimental facility2007In: Proceedings of the 15th International Conference on Nuclear Engineering (ICONE15), 2007Conference paper (Refereed)
  • 31.
    Karbojian, Aram
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    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.
    A scoping study of debris bed formation in the DEFOR test facility2009In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, no 9, p. 1653-1659Article in journal (Refereed)
    Abstract [en]

    Motivated to understand the processes which govern the formation and characteristics of a debris bed and hence its coolability during a postulated severe accident of a light water reactor, a new research program called DEFOR (DEbris FORmation) was initiated at the Royal Institute of Technology (KTH). This paper presents results obtained in scoping experiments conducted during an initial phase of the DEFOR program. The DEFOR-E test campaign is concerned with the DEFOR test facility commissioning and exploratory study of phenomena occurred during a debris bed formation. Binary oxide mixtures at different superheat temperatures were used as the corium melt simulants. The scoping experiments revealed the effect of water pool depth and subcooling, melt mass and material properties on the debris bed characteristics. Insights gained from the DEFOR-E test campaign help guide the scaling, design and operation of the subsequent experiments in the DEFOR program.

  • 32.
    Kozlowski, Tomasz
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Roshan, Sean
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kubarev, Andrej
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Deterministic Safety Analysis Group. Review for the period of 2008 – 20102010Report (Other academic)
    Abstract [en]

    SSM has signed an agreement with KTH and Chalmers for a long-term commitment within

    the deterministic safety analysis (transient and severe accident analysis). The objective for

    this co-operation is to promote the national competence and to establish groups, whom can

    support SSM to perform safety analysis, reviews and inquiries, as well as participate in

    international projects and working groups within these areas.

    According to the agreement, the analysis groups at KTH and Chalmers have worked with

    R&D within the following areas:

    - plant analysis;

    - evaluation and contribution to international projects;

    - training and education of SSM's personnel.

    The first phase of the Technical Support Organization for the Deterministic Safety Analysis

    (TSO-DSA) agreement covered the period of 2008 ¡V 2010. The objective of this document is

    to describe the work performed within the TSO-DSA and summarize the TSO-DSA

    achievements for the initial 3-year period.

  • 33. Kozlowski, Tomasz
    et al.
    Wysocki, Aaron
    Gajev, Ivan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xu, Yunlin
    Downar, Thomas
    Ivanov, Kostadin
    Magedanz, Jeffrey
    Hardgrove, Matthew
    March-Leuba, Jose
    Hudson, Nathanael
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 67, p. 4-12Article in journal (Refereed)
    Abstract [en]

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one.

  • 34.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Davydov, Mikhail
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, T.N.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A study of ex-vessel debris formation in a LWR severe accident2007In: Proceedings of the International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work": Societe Francaise d'Energie Nucleaire, Curran Associates, Inc., 2007, p. 2848-2859Conference paper (Refereed)
    Abstract [en]

    In the paper we analyze phenomena that governdebrisformationand introduceacomprehensive framework to exhibit their interrelationship duringahypotheticalsevereaccidentinaboiling water reactor (BWR). We focus on phenomena feedbacks and identify key parameters which are believed to have significant effect ondebrispacking, including boiling regimes on fragments, their settling against steam flow stemming fromabottom bed. Based on scoping calculations for reactor scenarios, the prototypic range of the key parameters is delineated. Taking into account the practical and technical constraints of laboratory experiments with simulant fluids and results from calculations for experimental conditions, we establish feasibility and parameter ranges, under whichanew series of DEFOR-S "snap-shot" experiments shall be conducted to provide reactor relevant data and insights. Requirements on DEFOR-S experimental measurements and data analysis are also discussed in the paper.

  • 35.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Idaho National Laboratory, United States.
    An experimental study on debris formation with corium simulant materials2008In: International Conference on Advances in Nuclear Power Plants, ICAPP 2008, 2008, p. 1191-1199Conference paper (Refereed)
    Abstract [en]

    Characteristics of corium debris beds formed in a severe coremelt accident are studied in DEFOR-S experiments, in which binary-oxidic simulant-material melts are discharged into a water pool. Water subcooling and pool depth are found to significantly influence the fragment morphology and agglomeration. The DEFOR-S data to date are indicative of the effect of melt composition, notably eutectic vs. non-eutectic. Synthesis of the DEFOR-S observations with respect to debris fragment's shape and surface roughness suggest the governing role of competitive mechanism: drop breakup, freezing and solid fracture.

  • 36.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on debris formation with corium stimulant materials2008In: Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP 2008), American Nuclear Society, 2008Conference paper (Refereed)
  • 37.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    THE DEFOR-S EXPERIMENTAL STUDY OF DEBRIS FORMATION WITH CORIUM SIMULANT MATERIALS2010In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 170, no 1, p. 219-230Article in journal (Refereed)
    Abstract [en]

    Characteristics of corium debris beds formed in a severe core melt accident are studied in the Debris Bed Formation-Snapshot (DEFOR-S) test campaign, in which superheated binary-oxidic melts (both eutectic and non-eutectic compositions) as the corium simulants are discharged into a water pool. Water subcooling and pool depth are found to significantly influence the debris fragments' morphology and agglomeration. When particle agglomeration is absent, the tests produced debris beds with porosity of similar to 60 to 70%. This porosity is significantly higher than the similar to 40% porosity broadly used in contemporary analysis of corium debris coolability in light water reactor severe accidents. The impact of debris formation on corium coolability is further complicated by debris fragments' sharp edges, roughened surfaces, and cavities that are partially or fully encapsulated within the debris fragments. These observations are made consistently in both the DEFOR-S experiments and other tests with prototypic and simulant corium melts. Synthesis of the debris fragments from the DEFOR-S tests conducted under different melt and coolant conditions reveal trends in particle size, particle sphericity, surface roughness, sharp edges, and internal porosity as functions of water subcooling and melt composition. Qualitative analysis and discussion reaffirm the complex interplay between contributing processes (droplet interfacial instability and breakup, droplet cooling and solidification, cavity formation and solid fracture) on particle morphology and, consequently, on the characteristics of the debris beds.

  • 38.
    Li, Liang Xing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, H. X.
    Chen, T. K.
    Ma, Wei Min
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental investigation on the dynamic characteristics of molten droplets and high-temperature particles falling in coolant2010In: 6TH INTERNATIONAL SYMPOSIUM ON MULTIPHASE FLOW, HEAT MASS TRANSFER AND ENERGY CONVERSION / [ed] Guo LJ; Joseph DD; Matsumoto Y; Sommerfeld M; Wang YS, 2010, Vol. 1207, p. 292-299Conference paper (Refereed)
    Abstract [en]

    This paper presents the dynamic characteristics of molten droplets and hot particles at the very beginning of their falling into coolant pools, which are of importance to the subsequent interactions such as fragmentation of the droplets in coolants. The falling course of a single droplet or a single hot particle was recorded by a high-speed camera and a curve of velocity vs. time was obtained. Emphasis was placed on the effects of the droplet's size and temperature, the coolant's temperature and properties, and the droplet's physical properties on the moving behavior. Tests with hot particles were also performed for a comparison with the droplets. The results for the all cases showed that the velocity of a falling droplet/particle decreased rapidly but rebounded shortly, at the beginning of droplet/particle falling in the coolant. Following such a V-shaped evolution in velocity, the droplet/particle slows down gradually till a comparatively steady velocity. An increase in either coolant temperature or droplet temperature results in a larger velocity variation in the "J-region", but a smaller deceleration after it moves out of the "J-region". The elevated volatility of a coolant leads to a steeper deceleration in the "J-region" and beyond. The bigger size of a particle leads to a greater velocity variation in the "J-region" and terminal velocity. A high melting point and thermal conductivity as well as lower heat capacity contribute to dramatic variation in the "J-region" and low terminal velocity.

  • 39.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gong, Shengjie
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental Study of Two-Phase Flow Regime and Pressure Drop in a Particulate Bed Packed with Multidiameter Particles2012In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 177, no 1, p. 107-118Article in journal (Refereed)
    Abstract [en]

    This paper documents an experimental study on two-phase flow regimes and frictional pressure drop characteristics in a particulate (porous) bed packed with multidiameter (1.5-, 3-, and 6-mm) glass spheres. The experimental results provide new data to validate/develop hydrodynamic models for coolability analysis of debris beds formed in fuel-coolant interactions during a postulated severe accident. The POMECO-FL test facility is employed to perform the experiment, with the spheres packed in a test section of 90 mm diameter and 635 mm height. The pressure drops are measured for air/water two-phase flow through the packed bed, and flow patterns are obtained by means of visual observations. Meanwhile, local void fraction in the center of the bed is measured by a microconductive probe.The experimental results show that the frictional pressure drop of single-phase flow through the bed can be predicted by the Ergun equation, if the area mean diameter of the particles is chosen in the calculation. Given the so-determined effective particle diameter, the estimation of the Reed model for two-phase flow pressure gradient in the bed has a good agreement with the experimental data. The characteristics of the local void fraction can be used to predict flow pattern and mean void fraction. It is observed that slug flow prevails when the mean void fraction is <0.5, whereas annular flow dominates after the mean void fraction is >0.7. If the effective particle diameter is further used as an influential parameter in flow pattern identification, the observed flow regimes of two-phase flow in porous media are well predicted by the existing flow pattern map.

  • 40.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Experimental Study on Dryout Heat Flux in Particulate Bed Packed with Irregular Particles2011In: Proceedings of the 2011 international congress on advances in nuclear power plants: ICAPP2011, American Nuclear Society, 2011Conference paper (Refereed)
    Abstract [en]

    This paper is concerned with reducing uncertainty in quantification of debris bed coolability in ahypothetical severe accident of light water reactors (LWRs), when the molten corium is relocatedinto a water pool, fragmenting and forming a particulate debris bed on the pool bottom.The test facility POMECO-HT at KTH is employed to investigate the coolability of particulatebeds which have some characteristics of a prototypical debris bed, such as packed with particlesof multiple sizes and irregular shapes. The facility features a high power capacity (up to 2.1MW/m2) which enables to obtain the dryout heat flux of a particulate bed formed by the DEFORparticles [1-2] for top-flooding and bottom-injection schemes. The particulate beds are chosen insuch a way that some characteristics of debris bed prototypicality analyzed in our previous study[3] can be reflected, so that the data can be used to confirm the previous analysis results. Threeparticulate beds, named Bed-1 through Bed-3, were employed in the present study to obtain thecorresponding dryout heat fluxes. Bed-1 and Bed-2 are both packed with sand particles whosesize distributions are similar to that of the DEFOR debris particles, with Bed-2 having a simulantcake embedded in the upper layer of the bed. Bed-3 is packed with DEFOR debris particles.The experimental results show that the dryout heat flux of the top-flooding beds can be predictedb y the Reed’s model. The bottom injection of coolant increases the dryout heat flux significantly,i.e., with an increase in water injection flowrate, the value of dryout heat flux is elevated.Meanwhile, the elevation of dryout position is moving upwards with increasing bottom-injectionflowrate. When a cake exits in a particulate bed, the dryout heat flux of the bed is significantlydecreased (up to 50%), and the dryout points a l w a y s located in the cake, for both the topflooding and bottom-fed configurations.

  • 41.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Huixiong
    State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University.
    Chen, Tingkuan
    State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dynamic characteristics of molten droplets and hot particles falling in liquid pool2010In: Frontiers of Energy and Power Engineering in China, ISSN 1673-7393, Vol. 4, no 2, p. 246-251Article in journal (Refereed)
    Abstract [en]

    The dynamic characteristics of molten droplets and hot particles at the very beginning of their fall into coolant pools are presented. The falling course of a single droplet or a single hot particle was recorded by a high-speed camera and a curve of velocity vs. time was obtained. Emphasis was placed on the effects of the droplet's size and temperature, the coolant's temperature and properties, and the droplet's physical properties on the moving behavior. The results for the all cases showed that the velocity of a falling droplet/particle decreased rapidly but rebounded shortly, at the beginning of droplet/particle falling in the coolant. Following such a V-shaped evolution in velocity, the droplet/particle slows down gradually to a comparatively steady velocity. An increase in either coolant temperature or droplet temperature results in a larger velocity variation in the "J-region", but a smaller deceleration when it moves out of the "J-region". The elevated volatility of a coolant leads to a steeper deceleration in the "J-region" and beyond. The bigger size of a particle leads to a greater velocity variation in the "J-region" and terminal velocity. A high melting point and thermal conductivity as well as lower heat capacity contribute to dramatic variation in the "J-region" and low terminal velocity.

  • 42.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on two-phase flow regimes and frictional pressure drop in porous media packed with multidiameter spheres2010In: Proceedings of 7th international conference on multiphase flow (ICMF2010), 2010Conference paper (Refereed)
  • 43.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental characterization of the effective particle diameter of a particulate bed packed with multi-diameter spheres2011In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 241, no 5, p. 1736-1745Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with uncertainty reduction in coolability analysis of a debris bed formed in fuel coolant interactions (FCI) during a postulated severe accident of LWRs. A test facility named POMECO-FL was designed and set up to investigate the friction laws of adiabatic single and two-phase flow through particulate beds which have the characteristics of the prototypical debris bed, such as packed with particles of multiple sizes or irregular shapes. The emphasis of the present study is placed on quantification of effective particle diameter of a particulate bed composed of multi-diameter spheres. Pressure drops are measured for water/air flow through the particulate beds packed with various combinations of spheres, and the effective particle diameters of the beds are obtained based on the pressure gradients and the Ergun equation. The results show that at low flowrate (Re < 7) the effective particle diameters can be represented by the area mean diameters of the particles in the beds, while at high velocity (Re > 7)the effective particle diameters are closer to the length mean diameters. If the area mean diameters are chosen as the effective particle diameters, the frictional pressure drops of two-phase flow in the beds can be predicted by the Reed model with good agreements.

  • 44.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental Investigations on Friction Laws and Dryout Heat Flux of Particulate Bed Packed with Multi-Size Spheres or Irregular Particles2011In: Proceedings of 19th International Conference on Nuclear Engineering (ICONE19), 2011Conference paper (Refereed)
  • 45.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental Study on the Effective Particle Diameter of a Packed Bed with Non-Spherical Particles2011Conference paper (Refereed)
  • 46.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental Study on the Effective Particle Diameter of a Packed Bed with Non-Spherical Particles2011In: Transport in Porous Media, ISSN 0169-3913, E-ISSN 1573-1634, Vol. 89, no 1, p. 35-48Article in journal (Refereed)
    Abstract [en]

    An experimental study is conducted to determine the characteristics of frictional pressure drops of fluid flow in porous beds packed with non-spherical particles. The objective is to examine the applicability of the Ergun equation to flow resistance assessment for packed beds with non-spherical particles. The experiments are carried out on the POMECOFL facility at KTH. Hollow spheres and cylinders are used to pack the beds. Either water or air is chosen as the working fluid. The experimental data show that the Ergun equation is applicable to all the test beds if the effective particle diameter used in the equation is chosen as the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles in each bed.

  • 47.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Toward Quantification of Debris Bed Coolability in Corium Risk Assessment2011In: 2011 ANS Winter Meeting and Nuclear Technology Expo, 2011, p. 1060-1061Conference paper (Refereed)
  • 48.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on pressure drop and dryout heat flux of two-phase flow in packed beds of multi-sized and irregular particles2012In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 242, p. 369-378Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with debris bed coolability in a postulated severe accident of light water reactors, where the debris particles are irregular and multi-sized. To obtain and verify the friction laws predicting the hydrodynamics of the debris beds, the drag characteristics of air/water single- and two-phase flow in a particulate bed packed with multi-sized spheres or irregular sand particles were investigated on the POMECO-FL test facility. The same types of particles were then loaded in the test section of the POMECO-HT facility to obtain the dryout heat fluxes of the particulate beds heated volumetrically. The effective (mean) particle diameter is 2.25 mm for the multi-sized spheres and 1.75 mm for the sand particles, determined from the Ergun equation and the measured pressure drop of single-phase flow through the packed bed. Given the effective particle diameter, both the pressure drop and the dryout heat flux of two-phase flow through the bed can be predicted by the Reed model. The experiment also shows that the bottom injection of coolant improves the dryout heat flux significantly and the first dryout position is moving upward with increasing bottom injection flowrate. Compared with top-flooding case, the dryout heat flux of the bed can be doubled if the superficial velocity of coolant injection is 0.21–0.27 mm/s. The experimental data provides insights for interpretation of debris bed coolability (how to deal with the multi-sized irregular particles), as well as high-quality data for validation of the coolability analysis models and codes.

  • 49.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Experimental Study on Two-Phase Flow and Coolability of Particulate Beds Packed with Multi-Size Particles2011In: Proceedings of NURETH-14 The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics, 2011Conference paper (Refereed)
  • 50.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Chapter 2.6 of the book "Nuclear Safety in Light Water Reactors: Severe Accident Phenomenology"2012In: Nuclear Safety in Light Water Reactors: Severe Accident Phenomenology / [ed] Bal Raj Sehgal, Academic Press, 2012, 1Chapter in book (Other academic)
12 1 - 50 of 96
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