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  • 1. Arzhanov, Vasily
    et al.
    Pazsit, I.
    Diagnostics of core barrel vibrations by in-core and ex-core neutron noise2003In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 43, no 04-jan, p. 151-158Article in journal (Refereed)
    Abstract [en]

    Diagnostics of core-barrel vibrations has-traditionally been made by use of ex-vessel neutron detector signals. We suggest that in Addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective diagnostics, especially for shell-mode vibrations. To this order, the theory of in-core noise induced by a fluctuating core boundary has been elaborated and applied to the diagnostics of beam and shell mode vibrations. The formulas were tested on some measurements taken in the Ringhals PWRs. The results confirm the validity of the model itself, and the possibilities for enhanced diagnostics were demonstrated. A more effective use of these novel possibilities requires more in-core detectors and/or better detector positioning.

  • 2.
    Bakardjieva, S.
    et al.
    Institute of Inorganic Chemistry, Czech Acad. Sci., Rez, Czech Republic.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Brissoneau, L.
    CEA, DEN, Cadarache, F-13108 St Paul lez Durance, France.
    Cheynet, B.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Fischer, E.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Journeau, C.
    CEA, DEN, Cadarache.
    Kiselova, M.
    Nuclear Research Institute UJV, Rez, 250 68, Czech Republic.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Wiss, T.
    European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, Germany.
    Improvement of the European thermodynamic database NUCLEA2010In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, no 1, p. 84-96Article in journal (Refereed)
    Abstract [en]

    Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is the European reference tool to achieve this goal. This database has been improved thanks to the analysis of bibliographical data and to EU-funded experiments performed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round-robin exercise has been launched in which a UO2-containing corium-concrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.

  • 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.
    Dinh, Truc-Nam
    et al.
    KTH, Superseded Departments, Physics.
    Konovalikhin, M. J.
    Sehgal, B. R.
    Core melt spreading on a reactor containment floor2000In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 36, no 4, p. 405-468Article in journal (Refereed)
    Abstract [en]

    The ex-vessel core melt spreading, cooling and stabilization is proposed for a nuclear power plant containment design. Clearly, the retention and coolability of the decay-heated core debris is very much the focal point in the proposed new and advanced designs so that, in the postulated event of a severe accident, the containment integrity is maintained and the risk of radioactivity releases is eliminated. The work reported here includes three tasks (i) to review related methodology and data base, (ii) to develop the scaling methodology and (iii) to validate the assessment methodology developed by the authors. The study is based on state-of-the-art knowledge of the melt spreading phenomenology, in particular, and, of severe accident phenomenology in general.

  • 6.
    Gudowski, Waclaw
    et al.
    KTH, Superseded Departments, Physics.
    Arzhanov, Vasily
    Broeders, C.
    Broeders, I.
    Cetnar, J.
    Cummings, R.
    Ericsson, M.
    Fogelberg, B.
    Gaudard, C.
    Koning, A.
    Landeyro, P.
    Magill, J.
    Pazsit, I.
    Peerani, P.
    Phlippen, P.
    Piontek, M.
    Ramstrom, E.
    Ravetto, P.
    Ritter, G.
    Shubin, Y.
    Soubiale, S.
    Toccoli, C.
    Valade, M.
    Wallenius, Janne
    KTH, Superseded Departments, Physics.
    Youinou, G.
    Review of the European project - Impact of Accelerator-Based Technologies on Nuclear Fission Safety (IABAT)2001In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 38, no 1-2, p. 135-151Article in journal (Refereed)
    Abstract [en]

    The IABAT project - Impact of Accelerator Based Technologies on Nuclear Fission Safety - started in 1996 in the frame of 4(th) Framework Programme of the European Union, R&D specific programme Nuclear fission safety 1994-1998, area A.2 Exploring innovative approaches/Fuel cycle concepts, as one of the first common European activities in ADS. The project was completed October 31, 1999. The overall objective of the IABAT project has been a preliminary assessment of the potential of Accelerator-Driven Systems (ADS) for transmutation of nuclear waste and for nuclear energy production with minimum waste generation. Moreover, more specific topics related to nuclear data and code development for ADS have been studied in more detail. Four ADSs have been studied for different fuel/coolant combinations: liquid metal coolant and solid fuel, liquid metal coolant and dispersed fuel, and fast and thermal molten salt systems. Target studies comprised multiple target solutions and radiation damage problems in a target environment. In a tool development part of the project a methodology of subcriticality monitoring has been developed based on Feynman-alpha and Rossi-alpha methods. Moreover, a new Monte-Carlo burnup code taking full advantage of continuous neutron cross-section data has been developed and benchmarked. Impact on the risk from high-level waste repositories fi om radiotoxicity reduction using ADS has been assessed giving no crystal-clear benefits of ADS for repository radiotoxicity reduction but concluding some important prerequisites for effective transmutation. In proliferation studies important differences between critical reactors and ADS have been underlined and non-proliferation measures have been proposed. In assessment of accelerator technology costing models have been created that allow the circular and linear accelerator options to be compared and the effect of parameter variations examined. The calculations reported show that cyclotron systems would be more economical, due mainly to the advantage of the cost of RF power supplies. However, the accelerator community regards with skepticism the possibility of transporting and extracting more than a 10mA beam current from a 1GeV cyclotron and therefore technical factors may limit the application of cyclotrons. Finally, this review summarizes development of nuclear data in the energy region between 20 Mev and 150 MeV. Neutron and proton transport data files for Fe, Ni, Pb, Th, U-238 and Pu-239 have been created. The high-energy part of the data files consists completely of results from model calculations, which are benchmarked against the available experimental data. Although there is obviously future work left regarding fine-tuning of several parts of the data files, the representation of nuclear reaction information up to 150 MeV is already better than can be attained with intranuclear cascade codes.

  • 7. Maschek, W.
    et al.
    Chen, X.
    Delage, F.
    Femandez-Carretero, A.
    Haas, D.
    Boccaccini, C. Matzerath
    Rineiski, A.
    Smith, P.
    Sobolev, V.
    Thetford, R.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Accelerator driven systems for transmutation: Fuel development, design and safety2008In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 50, no 2-6, p. 333-340Article in journal (Refereed)
    Abstract [en]

    European R&D for ADS design and fuel development is driven in the 6th FP of the EU by the EUROTRANS Programme. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT The XT-ADS is designed to provide the experimental demonstration of transmutation. The EFIT, the European Facility for Industrial Transmutation, aims at a conceptual design of a full transmuter. A key R&D issue is the choice of an adequate fuel. Various fuel forms have been assessed and CERCER and CERMET fuels, specifically the matrices MgO and Mo, have finally been selected. Within EUROTRANS, the domain 'AFTRA' is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel database for the EFIT. The EFIT is optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. In the current paper the fuels under investigation are described, including their production route and their safety limits. First core designs of CERCER and CERMET fuelled 400 MWth EFITs have been developed within AFTRA. The trends found in the design studies and first safety analyses are presented.

  • 8. Nourgaliev, R. R.
    et al.
    Dinh, Truc-Nam
    KTH, Superseded Departments, Physics.
    Dinh, A. T.
    Haraldsson, H. O.
    Sehgal, B. R.
    The multiphase Eulerian-Lagrangian transport (MELT-3D) approach for modeling of multiphase mixing in fragmentation processes2003In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 42, no 2, p. 123-157Article in journal (Refereed)
    Abstract [en]

    A new numerical approach for modeling of multiphase mixing during melt jet/droplet fragmentation process is developed. Melt or debris movements are simulated by a particle transport model in a Lagrangian formulation, while thermohydraulic conditions of the surrounding medium are obtained from solution of the Navier-Stokes and energy-conservation equations written in an Eulerian formulation. The Lagrangian and the Eulerian solutions are coupled and advanced in time, with source terms included to model the interactions between the particle and the continuum phases. The method is validated against isothermal solid-sphere, and drop fragmentation experiments. It is found that the model is capable of describing the evolution of the melt-coolant multiphase mixing process with reasonable accuracy. The method is then applied to investigate fragmentation of a continuous jet. Effects of variations in jet/coolant velocities, and of coolant thermophysical properties are analyzed, with particular emphasis on their implications for the fragmentation and mixing processes.

  • 9.
    Sehgal, B. R.
    et al.
    KTH. KTH, Stockholm, Sweden..
    Van Dorsselaere, J. P.
    Albiol, T.
    Jacquemain, D.
    Journeau, C.
    Major achievements after 4,5 years of SARNET (Severe Accident Research Network of Excellence)2010In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, no 1, p. 1-1Article in journal (Other academic)
  • 10.
    Tesinsky, Milan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Impact of reflector on Doppler feedback in fast reactorsIn: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224Article in journal (Other academic)
  • 11.
    Tran, Chi-Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation2009In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, no 8, p. 849-859Article in journal (Refereed)
    Abstract [en]

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

  • 12.
    Tran, Chi-Thanh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application2009In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, no 8, p. 860-871Article in journal (Refereed)
    Abstract [en]

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

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