kth.sePublications
Change search
Refine search result
1234 1 - 50 of 178
CiteExportLink to result list
Permanent link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf
Rows per page
  • 5
  • 10
  • 20
  • 50
  • 100
  • 250
Sort
  • Standard (Relevance)
  • Author A-Ö
  • Author Ö-A
  • Title A-Ö
  • Title Ö-A
  • Publication type A-Ö
  • Publication type Ö-A
  • Issued (Oldest first)
  • Issued (Newest first)
  • Created (Oldest first)
  • Created (Newest first)
  • Last updated (Oldest first)
  • Last updated (Newest first)
  • Disputation date (earliest first)
  • Disputation date (latest first)
  • Standard (Relevance)
  • Author A-Ö
  • Author Ö-A
  • Title A-Ö
  • Title Ö-A
  • Publication type A-Ö
  • Publication type Ö-A
  • Issued (Oldest first)
  • Issued (Newest first)
  • Created (Oldest first)
  • Created (Newest first)
  • Last updated (Oldest first)
  • Last updated (Newest first)
  • Disputation date (earliest first)
  • Disputation date (latest first)
Select
The maximal number of hits you can export is 250. When you want to export more records please use the Create feeds function.
  • 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.
    Bechta, Sevostian
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Miassoedov, Alexei
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Journeau, Christophe
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Okamoto, Koji
    Univ Tokyo, Tokyo, Japan..
    Manara, Dario
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Bottomley, David
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Kurata, Masaki
    JAEA CLADS Lab, Iwaki, Fukushima, Japan..
    Sehgal, Bal Raj
    Stuckert, Jun
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Steinbrueck, Martin
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Fluhrer, Beatrix
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Keim, Torsten
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Fischer, Manfred
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Langrock, Gert
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Piluso, Pascal
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Hozer, Zoltan
    MTA EK, Budapest, Hungary..
    Kiselova, Monika
    UJV REZ As, Hlavni 130, F-25068 Husinec Rez, Czech Republic..
    Belloni, Francesco
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    Schyns, Marc
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    On the EU-Japan roadmap for experimental research on corium behavior2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed)
    Abstract [en]

    A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.

  • 3. 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.

  • 4. 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.

  • 5.
    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.
    Development and application of a surrogate model for quick estimation of ex-vessel debris bed coolability2020In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 370, article id 110898Article in journal (Refereed)
    Abstract [en]

    During a hypothetical severe accident of a Nordic boiling water reactor (BWR), an ex-vessel particulate debris bed is expected to form in the flooded lower drywell due to melt-coolant interactions after vessel failure. The key parameter to evaluate debris bed coolability is the dryout heat flux (DHF) or dryout power density, representing the limit of heat removal capacity by the coolant. Several numerical codes such as COCOMO have been developed to simulate thermal hydraulics in multi-dimensional debris beds and predict the cooling limit, but they are computationally expensive and not suitable for probabilistic risk analysis. This paper aims to develop a surrogate model which can serve as a quick-estimate tool for the dryout power density of a heap-like debris bed in a saturated water pool. The dryout power density predicted from the COCOMO code is treated as the full model. A characteristic factor is introduced as the dryout power density ratio between the multi-dimensional debris bed (predicted by COCOMO code) and the corresponding one-dimensional debris bed (predicted by Lipinski 0-D model). The characteristic factor is correlated by the Kriging method with six parameters: bed porosity, particle diameter, debris mass, bed slope, cavity radius and containment pressure. After the surrogate model is trained and validated, it is employed to analyze the coolability of prototypical debris beds of a reference Nordic BWR, given the bed mass and containment pressure from MELCOR simulation. Coolability maps are produced as quick look-up diagrams for identification of coolable domain with the variation of porosity, particle diameter and slope angle. A preliminary uncertainty analysis is performed to demonstrate the effect of uncertain input parameters on non-coolable domain.

  • 6.
    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.
    Development of surrogate model for debris bed coolability analysis2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6770-6779Conference paper (Refereed)
    Abstract [en]

    The cornerstone of severe accident management (SAM) strategy of a Nordic boiling water reactor (BWR) is to flood the reactor cavity with water from the pressure suppression pool before failure of the reactor pressure vessel (RPV). The idea is to form a deep water pool which can accommodate the corium ejected from the RPV breach and cool the debris bed in the reactor cavity. Hence, assessment of debris bed coolant in the deep water pool is of paramount importance to the qualification of this SAM strategy. For the coolability analysis of a debris bed, one needs to estimate the dryout heat flux/power density of the particle bed, which is considered as the limit for heat removal capacity of coolant. For a multi-dimensional debris bed, the dryout power density can be assessed only by numerical simulation of two-phase flow and heat transfer in porous media. Since the numerical simulation is computationally expensive, it is neither suitable for massive calculations, nor feasible to be implemented into a system code (e.g. MELCOR). There is a clear need to develop a fast-running tool to estimate the dryout power density of a prototypical debris bed. The present study is concerned with development of a surrogate model which is sufficient for PSA study or capable of coupling with the MELCOR code without significant sacrifice of computational efficiency. The surrogate model is conceived from the coolability database predicted by COCOMO which is a mechanistic code for simulating thermal-hydraulic response of debris bed and has been extensively validated and applied in our previous studies [1][2]. The comparative results show that the surrogate model is not only able to predict the coolability limit of a debris bed, but also employed in the sensitivity study of bed’s characteristics (e.g., particle diameter, bed geometry and porosity) and the uncertainty and risk analysis.

  • 7.
    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.

  • 8.
    Chen, Yangli
    et al.
    KTH.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 143, article id 107490Article in journal (Refereed)
    Abstract [en]

    The Best Estimate Plus Uncertainty approach requires the knowledge of input uncertainties for the uncertainty propagation with best-estimate codes. Inaccurate judgement of some model parameter uncertainties related to the dominant physical phenomena may result in misestimation of the safety margin. This paper presents a framework of inverse uncertainty quantification (UQ) to quantify model parameter uncertainties in order to address this issue. It is applied to TRACE simulation of a large break loss of coolant accident conducted on the FIX-II facility, and peak cladding temperature (PCT) is the simulation output. Sensitivity analysis identifies the parameters of the critical flow model as the most influential to the PCT. The inverse UQ is performed based on Bayesian framework, which adopts Markov Chain Monte Carlo sampling and surrogate modelling algorithms. The quantified uncertainties of the model parameters are the desired results from the inverse UQ process, which are useful in BEPU studies.

  • 9.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zhang, Huimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds2022In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 165, article id 108643Article in journal (Refereed)
    Abstract [en]

    The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-termcoolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bedformed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of theMELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debrisbed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integralanalysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debrisbed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on thebed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns.Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity poolsaturation and containment venting.

    Download full text (pdf)
    fulltext
  • 10.
    Chen, Yangli
    et al.
    KTH.
    Zhang, Huimin
    KTH.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed)
    Abstract [en]

    This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

  • 11.
    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.

  • 12. 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.

  • 13. 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.

  • 14.
    Deng, Yucheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guo, Qiang
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Fang, Di
    KTH.
    Xiang, Yan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A numerical study on the levitation system for droplet preparation in a fuel-coolant interaction experiment2023In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 159, article id 104639Article in journal (Refereed)
    Abstract [en]

    The MISTEE facility at KTH was designed to investigate the process and phenomena of a molten droplet falling into a water pool that may be encountered in fuel-coolant interactions (FCI) during a severe accident of light water reactors. An aerodynamic levitation mechanism is proposed to hold the molten droplet during its preparation (melting and heating up to a prescribed temperature) in an induction furnace. The crucible is flushed with argon through an injection nozzle at the bottom to prevent the droplet from falling out of the crucible. A numerical simulation of the aerodynamic levitation system is performed in the present study with the objective of determining and optimizing the design. The problem was simplified as an isothermal two-phase flow in an axisymmetric geometry. The simulation is realized through ANSYS Fluent v17 platform, which employs the VOF method to track interfaces between two phases and the SST k-omega model to describe turbulence flow of argon gas. The numerical model is validated against tests performed in the MISTEE facility after mesh sensitivity study. It is then applied to investigate the impacts of various parameters on the facility levitation capability and the droplet stability. According to the simulation results, stable molten droplets can be obtained in the designed experimental setup. The simulation also provides the appropriate values of argon inlet velocity and sample mass at which a stable droplet can be obtained inside the crucible before its discharge. Either higher or lower inlet velocity will destabilize the formation of the droplet. Considering the temperature-dependent melt properties, both surface tension and viscosity affect the movement and deformation of the molten droplet. The wettability of melt on the crucible wall is critical to droplet formation, and it is found that a poor wettability can ensure the levitation of droplet.

  • 15.
    Deng, Yucheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guo, Qiang
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Fang, Di
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xiang, Yan
    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 droplet quench and steam explosion in boric acid solutions2023In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 166, article id 104970Article in journal (Refereed)
    Abstract [en]

    Boric acid (H3BO3) is widely adopted as an additive in the coolant of light water reactors for reactivity control, but its effect on fuel coolant interactions (FCI) during severe accidents (especially on steam explosion) was rarely investigated. To examine the effect of the boric acid additive in coolant on steam explosion, a series of molten droplet-coolant interaction tests using H3BO3 solutions (with concentration ranging from 0-3.2% by weight) is carried out in the present study. The characteristics of melt-coolant interactions are the occurrence probability of typical phenomena (no fragmentation, minor fragmentation, or spontaneous steam explosion), lateral deformation ratio, quench depth, pressure impulse and debris particle size distribution. The statistical data of such characteristics are obtained through repeating 20 runs of the same test category. The experimental results show that the H3BO3 addition in coolant has various impacts on the above-mentioned characteristics of melt-coolant interactions, depending on the H3BO3 concentration. In particular, the probability of steam explosion sightly decreases as the H3BO3 concentration increases from zero to 1.2 wt.%, but significantly increases as the H3BO3 concentration further increases to 3.2 wt.% trough 2.2 wt.%. Namely, the inhibiting effect of boric acid on steam explosion is diminishing with increasing H3BO3 concentration beyond 1.2 wt.%. It is also found that both melt and coolant temperatures are crucial parameters impacting the likelihood and energetics of steam explosion.

  • 16.
    Deng, Yucheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guo, Qiang
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xiang, Yan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Fang, Heng
    KTH, School of Electrical Engineering and Computer Science (EECS), Intelligent systems, Robotics, Perception and Learning, RPL.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on the effect of chemical additives in coolant on steam explosion2024In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 218, article id 124818Article in journal (Refereed)
    Abstract [en]

    In assessment of severe accident risk in light water reactors (LWRs), steam explosion is a nonnegligible phenomenon following a relocation of core melt (corium) into coolant, and thus various research efforts have been paid to steam explosion. There had been numerous studies showing that the occurrence of steam explosions is influenced by several factors such as melt and coolant temperatures, melt materials, non-condensable gasses, etc. However, most of the existing experiments used deionized (DI) water or tap water as coolant, with little consideration of the effect of chemicals (e.g. boric acid, sodium hydroxide, sodium phosphate) commonly applied in reactor coolant. To examine the effect of the chemical additives in coolant on steam explosion, the present study performs a series of molten Tin droplet-coolant interaction tests using DI water and different chemical solutions, including H3BO3 solutions, NaOH + H3BO3 neutral solutions, and Na3PO4 + H3BO3 neutral solutions. The experimental results show that adding NaOH and Na3PO4 in boric acid solution significantly affects the occurrence probability of spontaneous steam explosion, because of the presence of PO43− and H+ ions. When different solutions have equivalent concentrations of H3BO3, the peak pressure values of the spontaneous steam explosion of Sn droplets are similar among various solutions. Compared with those in DI water, steam explosion in the chemical solutions occurs predominantly within a narrow range of depth from 28 mm to 40 mm and produces a much higher peak pressure. This implies that more energetic steam explosions may occur in the chemical solutions.

  • 17.
    Dong, Shichang
    et al.
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Gong, Shengjie
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Zhang, Botao
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Xiong, Zhenqin
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Yuan, Yidan
    China Nucl Power Engn Co Ltd, Beijing, Peoples R China..
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Theoretical model for subcooled upward flow boiling heat transfer and critical heat flux for an inclined downward heated surface2023In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 213, article id 124351Article in journal (Refereed)
    Abstract [en]

    The in-vessel retention system and ex-vessel retention system are very important to the safety of nu-clear power plants under severe accidents. While the success of such safety systems relies on well un-derstanding the corresponding physical mechanisms of boiling heat transfer and critical heat flux (CHF). Challenges till remain in accurately predicting the subcooled flow boiling curve especially in the low-pressure and low-flow conditions due to its complex boiling phenomenon. The present study introduces a theoretical model to predict the boiling curve and critical heat flux for subcooled flow boiling in in-clined downward heated rectangular channel. The proposed model well estimates the transition from forced convection, isolated bubble nucleate boiling to fully developed boiling regime by considering the growth and interaction of bubbles. Through probability analysis of bubbles' interaction, the proportion of heat flux in different boiling regimes is determined. In addition, the flow boiling CHF is predicted based on the probability analysis of dry spots. The new model is validated by the subcooled flow boil-ing experiments with vertical single-side heated channel under low-pressure and low-flow conditions. The predicted boiling curves are consistent with experimental results corresponding to different thermal-hydraulic parameters, such as pressure, mass flux, inlet subcooling and wall wettability (hydrophilic and hydrophobic), and the prediction error of CHF is within & PLUSMN;15%. Furthermore, the inclination effect on CHF is validated by the subcooled flow boiling experiments in inclined channel with the inclination angle varying from 0 & DEG; to 90 & DEG;, which shows the good applicability of the developed model. 

  • 18.
    Dong, Shichang
    et al.
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Gong, Shengjie
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Zhang, Botao
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Yuan, Yidan
    China Nucl Power Engn Co Ltd, Beijing, Peoples R China..
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Mechanistic critical heat flux model development for subcooled flow boiling based on superheated liquid sublayer depletion2022In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 153, p. 104445-, article id 104445Article in journal (Refereed)
    Abstract [en]

    Critical heat flux (CHF) refers to the limit of boiling transfer systems, and crossing this limit may jeopardize system safety. However, a clear understanding of the physical mechanisms of CHF is still lacking. In this study, a new CHF prediction model based on superheated sublayer depletion was established for subcooled flow boiling in an upward vertical tube at low pressure. The model is characterized by its developed determination of the superheated liquid sublayer thickness, net vapor generation location, forced convection heat transfer and liquid supplement caused by bubble turbulent fluctuations. The proposed CHF model was validated by a database covering the low pressure subcooled operational ranging over P = 0.1-2.15 MPa, G = 0.7-35 Mg/m2s, Delta Tin = 11-183.16 K, D = 0.7-12 mm, L/D = 4.2-115.55, and xeqout = -0.2673-0.0843. The model can accurately predict the trend of thermal-hydraulic and geometric factors' effects on the CHF. The prediction results have good prediction accuracy with an root-mean-square error (RMSE) of 15.21%, and overall error of +/- 25%. The proposed model also shows good adaptation to a non-water (refrigerant 113 and liquid nitrogen) system.

  • 19.
    Fang, Di
    et al.
    KTH.
    Xiang, Yan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Deng, Yucheng
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A numerical study of liquid film dynamics in multi-nozzle spray cooling of downward-facing surface2023In: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533, Vol. 161, article id 104383Article in journal (Refereed)
    Abstract [en]

    In a consideration of spray cooling as the potential cooling mechanism for the in-vessel melt retention (IVR) strategy of nuclear reactors because of its superior heat removal efficiency, the SPAYCOR experiment has been conducted at KTH to investigate the spray cooling capacity of multiple nozzles applied to a downward-facing heated surface. In the present study, the dynamics of liquid film on the downward-facing surface resulting from the multi-nozzle spray are numerically simulated by using a coupled Eulerian-Lagrangian method implemented in the OpenFOAM platform. Prior to simulation of the SPAYCOR experiment, the numerical approach is used to calculate two theoretical setups which have known analytical solutions, with the objective to validate the models in predicting liquid film dynamics either in spray or on an inclined surface. In the simulation of the SPAYCOR experiment, the predicted film morphology shows a good agreement with the experimental observation. What's more, the influential factors, including the inclination of the downward-facing heater surface, the nozzle-to-surface distance as well as the nozzle-array layout, are also investigated numerically in the present study. The simulation results show that a decreasing nozzle-to-surface distance does not only lead to a thicker liquid film and a lower velocity in the vicinity of each spray coverage, but also increases non-uniformity of the liquid film. The nozzles-array layout has little influence on the average liquid film thickness and velocity, but significantly affects the film morphology.

  • 20.
    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)
  • 21.
    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).

  • 22.
    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.

  • 23.
    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
    Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, United States.
    Uncertainty analysis of the OECD/NRC Oskarshamn-2 BWR stability Benchmark2014In: Proceedings of the International Conference on Physics of Reactors, PHYSOR 2014, Japan Atomic Energy Agency, JAEA , 2014Conference paper (Refereed)
    Abstract [en]

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. 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 (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. This paper shows also how an uncertainty method was implemented for the event. Comparing the calculated uncertainty with the measured uncertainty gives confidence in the BWR stability prediction.

  • 24. 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.

  • 25. Gong, S.
    et al.
    Mei, Y.
    Amin, M. Y.
    Zhang, B.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Orientation effect on heat transfer coefficient of a downward surface for flow boiling in a rectangular channel under low flow rate2020In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 153, article id 119594Article in journal (Refereed)
    Abstract [en]

    Natural convection boiling in channel with Arc-Shaped will be encountered in the IVR-ERVC (In-Vessel Retention measure by External Reactor Vessel Cooling) system in nuclear power plant under severe accident. The flow and heat transfer characters in this situation is simulated by flow boiling of deionized water in an inclined rectangular channel under low flow rates. This paper aims to separate various parameters (such as orientation, mass flow rate and inlet quality, etc.) to investigate their individual effects on heat transfer coefficient (HTC) in a rectangular channel with cross section of 17 mm × 10 mm. By using a preheater at the inlet of the rectangular channel, the inlet quality could be controlled and the two-phase flow situation could be observed before the fluids entering into the main heater region on one side of the channel wall in downstream. Thus the characteristics of HTC on the main heater could be investigated at different flow patterns. The channel orientations vary from 15 to 90°, the mass flow rates vary from 110 to 288 kg/(m2s) and the qualities vary from 0.003 to 0.036, respectively. Experimental results show that the mass flow rate and quality effects on the HTC are very weak in this study. However, the orientation angle effect on HTC shows an transition region within 45°~60°, while it slowly changes when the orientation angle is smaller than 45° and bigger than 60°. Such tendency could be well formulated by the error function. Compared with different empirical formulas of saturated boiling HTC, it is found that the Liu & Winterton correlation can well predict the experimental HTC results in 90° orientation channel. Based on such correlation and coupled with the error function, a new model was developed by considering the orientation effect, which has an error of ±15% comparing with the experimental data.

  • 26.
    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)
  • 27.
    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.

  • 28.
    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.

  • 29.
    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.

  • 30.
    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)
  • 31.
    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)
  • 32.
    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%).

  • 33. 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.

  • 34.
    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)
  • 35.
    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.

  • 36.
    Gong, Yaopeng
    et al.
    Tsinghua Univ, Dept Engn Phys, Beijing, Peoples R China.;China Nucl Power Engn Co Ltd CNPE, Beijing, Peoples R China..
    Zhang, Li
    China Nucl Power Engn Co Ltd CNPE, Beijing, Peoples R China..
    Yuan, Yidan
    China Nucl Power Engn Co Ltd CNPE, Beijing, Peoples R China..
    Guo, Qiang
    China Nucl Power Engn Co Ltd CNPE, Beijing, Peoples R China..
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Huang, Shanfang
    Tsinghua Univ, Dept Engn Phys, Beijing, Peoples R China..
    Density Measurement of Molten Drop With Aerodynamic Levitation and Laser Heating2022In: Frontiers in Energy Research, E-ISSN 2296-598X, Vol. 10, article id 892406Article in journal (Refereed)
    Abstract [en]

    Thermophysical properties of molten core materials (corium) are necessary input parameters of models and computer codes which predict the severe accident progression in light water reactors. The corium contains the components of UO2, ZrO2, Zr and Fe. The measurement of molten corium properties is a very challenging task due to high melting points of corium which can reach 3000 K. This paper presents a density measurement system for a molten drop based on techniques of aerodynamic levitation, laser heating and image processing. A sphere of alumina was firstly levitated by argon gas flow above a conical converging-diverging nozzle. The sphere was then heated up and melted into a liquid drop by a laser beam. The shape of the drop was recorded by a high-speed camera, and the density was calculated from image processing.

  • 37.
    Gong, Yaopeng
    et al.
    Department of Engineering Physics, Tsinghua University, Beijing, China; China Nuclear Power Engineering Co., Ltd (CNPE), Beijing, China.
    zhang, Li
    China Nuclear Power Engineering Co., Ltd (CNPE), Beijing, China.
    Yuan, Yidan
    China Nuclear Power Engineering Co., Ltd (CNPE), Beijing, China.
    Guo, Qiang
    China Nuclear Power Engineering Co., Ltd (CNPE), Beijing, China.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Huang, Shanfang
    Department of Engineering Physics, Tsinghua University, Beijing, China.
    Viscosity measurement of molten alumina and zirconia using aerodynamic levitation, laser heating and droplet oscillation techniques2023In: Heliyon, E-ISSN 2405-8440, Vol. 9, no 12, article id e22424Article in journal (Refereed)
    Abstract [en]

    Reliable thermophysical properties of core melt (corium) are essential for the accurate prediction of the severe accident progression in light water reactors. Zirconia is one of the most important materials in corium. Despite the high interest in the viscosity of molten zirconia, few experimental data have been reported due to its high melting temperature and high vapor pressure. In the present study, the viscosity of molten zirconia was measured using aerodynamic levitation, laser heating and droplet oscillation techniques. A material sample was levitated by argon gas flow in a conical nozzle and then melted into a droplet by laser beams. The initial quiescent droplet was forced to oscillate by the excitation of a loudspeaker, and the viscosity was deduced based on the characteristics of the droplet damped oscillation after the loudspeaker was turned off. The viscosity of molten alumina was first measured for verification of the measurement system. Afterwards the viscosity of molten zirconia was measured. The results showed that the viscosity of molten zirconia at melting temperature (2988K) was 12.87 ± 1.03 mPa s and decreased with increasing temperature. The measurement uncertainties are within 21 %.

  • 38.
    Gong, Yaopeng
    et al.
    Tsinghua University Beijing, China.
    Zhang, Li
    China Nuclear Power Engineering Co., Ltd Beijing, China.
    Yuan, Yidan
    China Nuclear Power Engineering Co., Ltd Beijing, China.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Huang, Shanfang
    Tsinghua University Beijing, China.
    Guo, Qiang
    China Nuclear Power Engineering Co., Ltd Beijing, China.
    Li, Chuanjun
    China Nuclear Power Engineering Co., Ltd Beijing, China.
    Liu, Yanping
    China Nuclear Power Engineering Co., Ltd Beijing, China.
    Surface Tension Measurement of Molten Zirconia with Aerodynamic Levitation and Laser Heating2023In: Proceedings of the 30th International Conference on Nuclear Engineering "Nuclear, Thermal, and Renewables: United to Provide Carbon Neutral Power", ICONE 2023, American Society of Mechanical Engineers (ASME) , 2023Conference paper (Refereed)
    Abstract [en]

    Thermophysical properties of corium are required in models and computer codes to predict the severe accident progression in light water reactors. However, the measurement of molten corium properties is challenging due to high melting points. This paper presents a surface tension measurement system for molten zirconia based on techniques of aerodynamic levitation and laser heating. Zirconia is one of the main components in corium and aerodynamic levitation is a contactless method to avoid interactions between the sample and container wall at high temperatures. A sample of zirconia was levitated by argon gas flow above a conical converging-diverging nozzle and then melted into a droplet by laser beams. The oscillation of molten zirconia was imaged by a high-speed camera. The resonant frequency was then obtained through image processing. Finally, the surface tension was derived according to the Rayleigh formula.

  • 39. 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.

  • 40.
    Guo, Qiang
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Deng, Yucheng
    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 quenching of metallic spheres in seawater2023In: Experimental Thermal and Fluid Science, ISSN 0894-1777, E-ISSN 1879-2286, Vol. 148, article id 110990Article in journal (Refereed)
    Abstract [en]

    Motivated by quenching of melt droplets and debris particles in seawater during a hypothetical severe accident of light water reactors (LWRs), the quenching process of stainless-steel spheres in a seawater pool was investigated in the present study. The polished spheres were pre-heated up to 1000celcius in an induction furnace with inserted atmosphere, and then immersed into the subcooled water pool in a chamber made of transparent quartz. A thermocouple was embedded in the center of the sphere to measure the history of the sphere's temperature, while a high-speed camera was employed to record the quenching process and vapor film dynamics. Quantitative data, e.g. film thickness and oscillation, of the vapor film evolution during the quenching process were obtained through an image processing program developed on the MATLAB platform.The experimental results indicated that the quenching rate was higher in seawater than in deionized water, and the vapor film collapsed at a temperature higher than the Leidenfrost temperature in deionized water. The trend appeared more significant with increasing subcooling of water. The comparison of the quenched spheres suggested the surface of the sphere in seawater achieved higher degree of discoloration and roughening than that in deionized water, probably due to the additives of salt which change water properties. The image processing and analysis revealed that the vapor film had different thickness profile along the upper and lower hemispheres, and the averaged film thickness was smaller in seawater than in de-ionized water during the stage of stable film boiling. The vapor film was thinning and oscillating with time, and its fluctuations appeared different frequencies and amplitudes at the upper and lower locations, which may explain the mechanism of the earlier collapse of vapor film in the quenching process of a high-temperature sphere in seawater.

  • 41.
    Guo, Qiang
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Manickam, Louis
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effects of salinity in coolant on steam explosion2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 4556-4567Conference paper (Refereed)
    Abstract [en]

    During a severe accident scenario of nuclear power plants, a steam explosion may occur when a substantial amount of molten core materials is rapidly ejected into a volatile coolant (water) pool, forming so-called Fuel-Coolant Interactions (FCI). The steam explosion poses a serious threat to the containment integrity. It is therefore important to understand and suppress the risk of steam explosions. The present study is concerned with mechanism of steam explosion on effect of coolant composition, i.e., to investigate how seawater impacts steam explosion energetics. For this purpose, a set of preliminary experiments were performed on spontaneous steam explosion by delivering a single molten tin droplet into a cold water pool at different levels of salinity on MISTEE facility. This paper presents the comparative results of the experimental data, including the influences of the salinity on probability of spontaneous explosion occurrence, explosion depth underwater and available thermal energy of droplet for explosion, as well as fragmentation (particle size distribution of debris % by mass). As the reference of the comparisons, the steam explosion characteristics from experiments in deionized water were employed. We experimentally observed that probability of spontaneous explosion occurrence increased in seawater, and more thermal energy of droplet was available for explosion when a droplet was self-triggered. The seawater at high salinity (35.16 g/kg) appeared remarkable enhancement on fragmentation. More experimental data are still needed to reveal more details and to develop a model for better understanding and prediction for the effects. The present data was helpful for prudential assessment on the seawater effects when it was employed as ultimate emergency cooling if NPPs located on sea coasts encounter Fukushima-like accidents.

  • 42.
    Guo, Qiang
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Manickam, Louis
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yu, Peng
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A design study on an aerodynamic levitation system for droplet preparation in steam explosion experiment2019In: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Conference paper (Refereed)
    Abstract [en]

    In order to investigate the mechanisms of steam explosion which may occur during a severe accident of light water reactors (LWRs), the MISTEE facility was developed at Royal Institute Technology (KTH) to visualize the micro interactions of steam explosion when a single molten droplet was falling into a water pool. For preparation of a molten droplet, an aerodynamic levitation system was proposed to prevent the droplet from falling out of the crucible during heating in an induction furnace by injecting argon gas through a purging line connected to the bottom nozzle of the crucible. To support the design of such levitation system, a numerical simulation of the aerodynamic levitation system was performed using the CFD code ANASYS FLUENT v16.2. The problem was simplified as adiabatic two-phase flow dynamics in a 2-D axisymmetric geometry. The VOF method is employed to track the interface of two phases (liquid metal and argon gas), and the SST k-omega model was chosen for turbulence. Various characteristics of droplet dynamics in incorporated with argon gas flowrates through the crucible were examined in the numerical simulation. The simulation results suggested there exists an optimal range of argon gas flowrate for levitating a coherent shape of droplet in the crucible. The wall adhesion had a considerable effect on initiating the levitation of the droplet, which means the properties of the inside surface of the crucible may play an important role in the levitation and discharge of the droplet. Proof-of-concept tests were carried out on the prototype of the design, and it was confirmed that the levitation system was able to fulfill its function, i.e., to keep the droplet in the crucible during its melting process, although the actual argon gas flowrates for levitation was higher than the predicted ones, probably due to the leakage of flow path and heat transfer which were not considered in the simulation. Generally speaking, the numerical simulation did not only help understand the hydrodynamic characteristics of the levitation system, but also provided insights on operational control and improvement of the system.

  • 43. 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.

  • 44.
    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.
    Numerical investigation on quench of an ex-vessel debris bed at prototypical scale2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 47-61Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.

  • 45.
    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.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubesManuscript (preprint) (Other academic)
  • 46.
    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.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, p. 189-202Article in journal (Refereed)
    Abstract [en]

    Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called "CRGT cooling"). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H-2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

  • 47.
    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.

  • 48.
    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 of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed)
    Abstract [en]

    In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.

  • 49.
    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.

  • 50.
    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.

1234 1 - 50 of 178
CiteExportLink to result list
Permanent link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf