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  • 1. Albiol, T.
    et al.
    Van Dorsselaere, J. P.
    Chaumont, B.
    Haste, T.
    Journeau, Christophe
    Meyer, Leonhard
    Sehgal, Bal Raj
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Schwinges, Bernd
    Beraha, David
    Annunziato, Alessandro
    Zeyen, Roland
    SARNET: Severe accident research network of excellence2010In: PROG NUCL ENERGY, 2010, Vol. 52, no 1, p. 2-10Conference paper (Refereed)
    Abstract [en]

    Fifty-one organisations network in SARNET (Severe Accident Research NETwork of Excellence) their research capacities in order to resolve the most important pending issues for enhancing, with regard to Severe Accidents (SA). the safety of existing and future Nuclear Power Plants (NPPs). This project. co-funded by the European Commission (EC) under the 6th Framework Programme, has been defined in order to optimise the use of the available means and to constitute sustainable research groups in the European Union. SARNET tackles the fragmentation that may exist between the different national R&D programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the organisations involved in SA research in Europe, plus Canada. To reach these objectives, all the organisations networked in SARNET contributed to a joint Programme of Activities, which consisted of: Implementation of an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents: Harmonization and re-orientation of the research programmes, and definition of new ones; Analysis of the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; Development of the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; Development of Scientific Databases in which all the results of research programmes are stored in a common format (DATANET); Development of a common methodology for Probabilistic Safety Assessment of NPPs; Development of short courses and writing a textbook on Severe Accidents for students and researchers; Promotion of personnel mobility amongst various European organisations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvement of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners. After this first period (2004-2008), co-funded by the EC, a further contract SARNET2 with the EC for the next four years started in April 2009 as part of the 7th Framework Programme. During this period, the networking activities will focus mainly on the remaining pending issues as determined during the first period, experimental activities will be directly included in the common work and the network will evolve toward complete self-sustainability. The bases for such an evolution are presented in the last part of the paper.

  • 2. Almyashev, V. I.
    et al.
    Granovsky, V. S.
    Khabensky, V. B.
    Krushinov, E. V.
    Sulatsky, A. A.
    Vitol, S. A.
    Gusarov, V. V.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Barrachin, M.
    Fichot, F.
    Bottomley, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effects during corium melt in-vessel retention2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, p. 389-399Article in journal (Refereed)
    Abstract [en]

    In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

  • 3. Alsmeyer, H
    et al.
    Albrecht, G
    Meyer, L
    Hafner, W
    Journeau, C
    Fischer, M
    Hellman, S
    Eddi, M
    Allelein, H J
    Burger, M
    Sehgal, Balraj
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Koch, M K
    Alkan, Z
    Petrov, J B
    Gaune-Escard, M
    Altstadt, E
    Bandini, G
    Ex-vessel core melt stabilization research (ECOSTAR)2005In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 235, no 2-4, p. 271-284Article in journal (Refereed)
    Abstract [en]

    The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical-chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO(2)-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.

  • 4.
    Amselem, Elias
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of Boiling Water Reactor Design and Operating Conditions Effect on Stability Behaviour2011Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    It is well known that boiling water reactors can experience inadvertent power oscillations. When such instability occurs the core can oscillate in two different modes (in phase mode and out of phase mode). In the late 90’s a stability benchmark was created using the stability data obtained from the experiments at the Swedish nuclear power plant of Ringhals-1. Data was collected from the cycles 14, 15 , 16 and 17. Later on, this data was used to validate the various models and codes with the aim of predicting the instability behavior of the core and understand the triggers of such oscillations. The current trend of increasing reactor power density and relying on natural circulation for core cooling may have consequences for the stability of modern BWR’s designs. The objective of this work is to find the most important parameters affecting the stability of the BWRs and propose alternative stability maps. For this purpose a TRACE/PARCS model of the Ringhals-1 NPP will be used. Afterwards a selection of possible parameters and dimensionless numbers will be made to study its effect on stability. Once those parameters are found they will be included in the stability maps to make them more accurate.

  • 5. Bakardjieva, S.
    et al.
    Barrachin, M.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Aleksandrov RIT/NITI, Russia.
    Bezdicka, P.
    Bottomley, D.
    Brissonneau, L.
    Cheynet, B.
    Dugne, O.
    Fischer, E.
    Fischer, M.
    Gusarov, V.
    Journeau, C.
    Khabensky, V.
    Kiselova, M.
    Manara, D.
    Piluso, P.
    Sheindlin, M.
    Tyrpekl, V.
    Wiss, T.
    Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET22014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 110-124Article in journal (Refereed)
    Abstract [en]

    In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident.

  • 6. Bandini, G.
    et al.
    Bubelis, E.
    Schikorr, M.
    Stempnievicz, M. H.
    Lázaro, A.
    Tucek, K.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Mansani, L.
    Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor2013Conference paper (Refereed)
    Abstract [en]

    The conceptual design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is under development within the LEADER project to meet the safety objectives of Gen IV nuclear energy systems. This paper presents the main results of the safety analysis for beyond design basis conditions, namely design extension conditions (DEC), which include the failure of prevention and mitigation systems, like the reactor scram in the so called unprotected transients. The main objective of this analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the ALFRED reactor. Several computer codes: SIM LFR, RELAP5, CATHARE, SPECTRA and TRACE are applied to evaluate the consequences of representative unprotected accident scenarios such as Loss of Flow, Loss of Heat Sink and Reactivity initiated accidents. Additionally, the consequences of steam generator tube rupture and partial sub assembly flow blockage events are assessed by means of appropriate fluid dynamic codes. 

  • 7. 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.

  • 8.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Particulate Debris Spreading and Coolability2017Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    In Nordic design of boiling water reactors, a deep water pool under the reactor vessel is employed for the core melt fragmentation and the long term cooling of decay heated corium debris in case of a severe accident. To assess the effectiveness of such accident management strategy the Risk-Oriented Accident Analysis Methodology has been proposed. The present work contributes to the further development of the methodology and is focused on the issue of ex-vessel debris coolability.

    The height and shape of the porous debris bed are among the most important factors that determine if the debris can be cooled by natural circulation of water. The bed geometry is formed in the process of melt release, fragmentation, sedimentation and packing of the debris in the pool. Bed shape is affected by the coolant flow that induces movement of particles in the pool and after settling on top of the bed. The later one is called debris bed self-leveling phenomenon.

    In this study, the self-leveling was investigated experimentally and analytically. Experiments were carried out in order to collect data necessary for the development of a numerical model with an empirical closure. The self-leveling model was coupled to a model for prediction of the debris bed dryout. Such coupled code allows to calculate the time necessary to have a coolable configuration of the bed. The influence of input parameters was assessed through sensitivity analysis in order to screen out the less influential parameters.

    Results of the risk analysis are reported as complementary cumulative distribution functions of the conditional containment failure probability (CCFP).

    Sensitivity analyses identified: effective particle diameter and debris bed porosity as the parameters that provide the largest contribution to the CCFP uncertainty. It is found that the effect of the initial maximum height of the bed on the CCFP is reduced by the self-leveling.

  • 9.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of scalable empirical closures for self-leveling of particulate debris bed2014In: Proceedings of ICAPP 201,  Paper 14330, American Nuclear Society, 2014, p. 14330-Conference paper (Refereed)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel is employed as a severe accident mitigation strategy in several designs of light water reactors. Geometrical configuration of the debris bed is one of the factors which define if the decay heat can be removed from the debris bed by natural circulation. A bed can be coolable if spread uniformly, while the same debris forming a tall mound-shape debris bed can be non-coolable. Two-phase flow inside the bed serves as a source of mechanical energy which can move debris, thus flatten and gradually reduce the height of the debris bed. There is a competition between the time scales for (i) reaching a coolable configuration of the bed by such “self-leveling” phenomenon, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local (i) gas velocity, and (ii) slope angle of the bed. The goal of this work is to obtain a dependency of particle motion rate on local slope angle and gas velocity expressed in non-dimensional variables, universal for particles of different shapes, sizes and materials. Such scaling approach is proposed in this work and validated against experimental data.

  • 10.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effectiveness of the debris bed self-leveling under severe accident conditions2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 95, p. 75-85Article in journal (Refereed)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under the reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. The success of such strategy is contingent upon the natural circulation effectiveness in removing the decay heat generated in the porous debris bed. The maximum height of the bed is one of the important factors which affect the debris coolability. The two-phase flow within the bed generates mechanical energy which can change the geometry of the debris bed by the "self-leveling" phenomenon. In this work.we developed an approach to modeling of the self-leveling phenomenon. Sensitivity analysis was carried out to rank the importance of the model uncertainties and uncertain input parameters i.e. the conditions of the accident scenario and the debris bed properties. The results provided some useful insights for further improvement of the model and reduction of the output uncertainties through separate-effect experimental studies. Finally, we assessed the self-leveling effectiveness, quantified its uncertainties in prototypic severe accident conditions and demonstrated that the effect of self-leveling phenomenon is robust with respect to the considered input uncertainties.

  • 11.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Empirical closures for particulate debris bed spreading induced by gas-liquid flow2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, p. 19-25Article in journal (Refereed)
    Abstract [en]

    Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called "self-leveling" phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different shapes and size distributions.

  • 12.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Preliminary Risk assessment of ex-vessle debris bed coolability for a Nordic BWRIn: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed)
    Abstract [en]

    In Nordic design of boiling water reactors (BWRs) a deep water pool under the reactor vessel is employed as a severe accident management strategy for the core melt fragmentation and the long term cooling of corium debris. The height and shape of the debris bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry is formed as a result of melt release, fragmentation, sedimentation and settlement on the containment basemat. After settlement, the shape can change with time due to movement of particles promoted by the coolant flow (debris bed self-leveling process). Both aleatory (accident scenario, stochastic) and epistemic (modeling, lack of knowledge) uncertainties are important for assessing the risks.

     

    The present work describes a preliminary risk analysis of debris bed coolability for Nordic BWRs under severe accident conditions. It was assumed that once debris remelting starts containment failure becomes imminent. Such assumption allows to estimate the containment failure probability by calculating the probability that the time necessary for the spreading debris bed to achieve a coolable configuration will be shorter than the onset time of debris bed re-melting. An artificial neural network was employed as a surrogate model (SM) for the mechanistic full model (FM) of the debris spreading in order to achieve computationally efficient propagation of uncertainties. The effect of uncertainty in the ranges and probability density functions (PDFs) of the input parameters was addressed. Parameters defining shapes of the PDFs were varied for three different distribution families (beta, truncated normal and triangular). The results of the risk analysis were reported as complementary cumulative distribution functions (CCDFs) of the conditional containment failure probability (CCFP). It is demonstrated that CCFP can vary in wide ranges depending on the randomly selected combinations of the PDFs of the input parameters. Given the selected ranges of the input parameters, sensitivity analyses identified: the effective particle diameter and the debris bed porosity as the largest contributors to the CCFP uncertainty. It was shown that the self-leveling phenomenon reduces sensitivity of debris coolability to the initial shape of the bed. However, the initial shape remains an important uncertainty factor for the most likely values of the particle size and porosity. Importance of the initial shape increases when the effectiveness of the self-leveling is small (e.g. in case of high initial temperature or heat up rate of the debris). Findings of this work in combination with consideration of the necessary efforts can be used for prioritization of the future research on obtaining new information on the uncertain parameters.

  • 13.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity and uncertainty analysis for predication of particulate debris bed self-leveling in prototypic Severe Accident (SA) conditions2014In: Proceedings of ICAPP 2014: Proceedings of ICAPP 2014, Paper 14329, American Nuclear Society, 2014, p. 14329-Conference paper (Refereed)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase flow inside the bed serve as a source of mechanical energy which can change the geometry of the debris bed by so called “self-leveling” phenomenon. The goals of this work are (i) to further develop self-leveling modeling approach and validate it against data produced in a new series of PDS-C (Particulate Debris Spreading Closures) experiments, and (ii) to carry out sensitivity-uncertainty analysis for the debris bed spreading for the selected cases of prototypic severe accident conditions. The model has been extended to predict spreading in both planar and axisymmetric geometries. The performed sensitivity analysis ranks the importance of different uncertain input parameters such as accident conditions, debris bed properties, modeling parameters and closures. The knowledge about the most influential parameters is important for further improvement of the model and for efficient reduction of output uncertainties through focused, separate-effect experimental studies. Finally, we report results for particulate debris spreading in prototypic severe accident scenarios with assessment of uncertainties.

  • 14.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yakush, S. E.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effect of self-leveling on debris bed coolability under severe accident conditions2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, p. 246-259Article in journal (Refereed)
    Abstract [en]

    Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWR5, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the initially non-coolable configurations, a significant portion becomes coolable due to debris bed self-leveling.

  • 15.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yakush, Sergey
    Institute for Problems in Mechanics, Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow, 119526, Russia.
    Kudinov, Pavel
    Validation of DECOSIM code against experiments on particle spreading by two-phase flows in water pool2016In: Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, NUTHOS-11, 2016, article id N11A0531Conference paper (Refereed)
    Abstract [en]

    Validation simulations by DECOSIM code are performed against recent PDS-P experiments on particle spreading in a planar vertical water pool with bottom air injection. The model implemented in the code considers two-fluid formulation (water, air), turbulence effects in liquid phase are taken into account by k-epsilon model with additional generation terms accounting for two-phase effects. Particles are described by Lagrangian model, with turbulent dispersion modeled by random-walk model. Simulations are performed in conditions corresponding to experimental setup, the test section was a plane rectangular tank of variable length (0.9 and 1.5 m) and pool depth (0.5, 0.7, and 0.9 m), the superficial gas injection velocity ranged between 0.12 and 0.69 m/s. Sedimentation of spherical stainless steel (1.5 and 3 mm) and glass (3 mm) particles was calculated and compared with experiments with respect to the mean spreading distance and lateral distributions of mass fraction of particles. Reasonable agreement between the results obtained and experimental measurements is achieved for all pool geometries, gas injection rates, and particle types, confirming adequacy of the modeling approach and suitability of DECOSIM code for severe accident analysis related to debris bed formation. Possible ways to further reduction of uncertainty in model validation are discussed.

  • 16.
    Beltran Arroyos, Guillem
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Investigation of Conditions for Activation of Rupture Disk in BWR Containment Filtering System2011Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    Due to the Three Mile Island accident in 1979 the Swedish government took the decision in 1986 to impose a pressure relief system for Swedish BWR’s which prevents containment overpressure in case of LOCA. This pressure relief system consists of a rupture disks in two different systems, non-filtered system 361 and filtered system 362.

    During a steam line break it is not clear if an unjustified activation of rupture disk 361 or 362 could possibly occur. If significant amount of nitrogen will leak out from the containment then, there is a risk of low pressure in the containment (e.g. due to activation of containment spray) with leaking rupture disks, which might cause air inflow to the containment and burning of hydrogen, so conditions of activation of rupture disk must be studied.

    The main objective of this master thesis is the investigation of conditions of activation of rupture disk in BWR containment filtering system. In order to find out these conditions specific software called GOTHIC has been used.

    The methodology of this master thesis has been modeling different containments with GOTHIC software; this thesis work will go from a simple GOTHIC model, that consist in nine lumped control volumes connected by flow paths, until a more complex GOTHIC model that consist in a combination of lumped and 3D control volumes, connected among them by flow paths and 3D connectors.

    A large LOCA in the upper part of the reactor vessel will be considerate, due to this severe accident; conditions for the activation of the rupture disk will be complying. It has to be mentioned that pressure in the lumped modeling will be lower than pressure in the 3D volumes. Activation time for the lumped modeling will be 8,5 seconds after the steam break for system 362 and activation time for 3D modeling will be 2,8 seconds for system 362 as well. In neither case 361 system will be activated.

    Considering this is a nuclear safety study and accuracy must be a key point, for further investigations it might be more than advisable using 3D control volumes instead of lumped control volumes.

    It has to be mentioned also that due to there is no experimental data, uncertainty regarding to the results exist, and if a further safety analysis want to be done, sensitive study of the parameters implemented on GOTHIC software should be performed in the future.

  • 17.
    Bertran Morancho, Joan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    TRACE Code Validation for Natural Circulation During Small Break LOCA in EPR-Type Reactor2011Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    The PWR PACTEL test facility was built in Lappeenranta (Finland) to gain experience in thermal-hydraulics behavior of vertical steam generators used by EPR (European Pressurized Water Reactor) during SBLOCA (Small Break Loss of Coolant Accident) transient, which involves natural circulation phenomenon. The benchmark, which consisted of blind and open part, offered a unique opportunity for code users to improve and test their knowledge and skills in developing the input deck models and performing calculations. For a purpose of this investigation, Royal Institute of Technology (KTH) has developed two TRACE code models.

    The main point of this thesis is to study TRACE code performance during SBLOCA transient and sensitivity of the developed TRACE models for the time and space convergence, which is very important for transients involving natural circulation phenomenon. Four different nodalizations coarse, inter-medium, fine and fine-sliced (space convergence), are designed for both designed models, which are calculated with different maximum time steps (time convergence). The results assessment was made by comparisons of the main parameters e.g.: Pressure of upper plenum, Inlet/outlet temperature of Core/SGs, Collapsed water level in the core, among others. In addition, discussion about vertical SGs performance during natural circulation phenomenon and conclusions for both, code users and developers, are provided.

  • 18.
    Bora Pekicten, Aziz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Assembly homogenization of light water reactors by a monte carlo reactor physics method and verification by a deterministic method2011Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
  • 19. Bottomley, D.
    et al.
    Stuckert, J.
    Hofmann, P.
    Tocheny, L.
    Hugon, M.
    Journeau, C.
    Clement, B.
    Weber, S.
    Guentay, S.
    Hozer, Z.
    Herranz, L.
    Schumm, A.
    Oriolo, F.
    Altstadt, E.
    Krause, M.
    Fischer, M.
    Khabensky, V. B.
    Bechta, Sevostian V.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Veshchunov, M. S.
    Palagin, A. V.
    Kiselev, A. E.
    Nalivaev, V. I.
    Goryachev, A. V.
    Zhdanov, V.
    Baklanov, V.
    Severe accident research in the core degradation area: An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center2012In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 252, p. 226-241Article in journal (Refereed)
    Abstract [en]

    The International Science and Technology Center (ISTC) was set up in Moscow to support non-proliferation of sensitive knowledge and technologies in biological, chemical and nuclear domains by engaging scientists in peaceful research programmes with a broad international cooperation. The paper has two following objectives: to describe the organization of complex, international, experimental and analytical research of material processes under extreme conditions similar to those of severe accidents in nuclear reactors and, to inform briefly about some results of these studies. The main forms of ISTC activity are Research Projects and Supporting Programs. In the Research Projects informal contact expert groups (CEGs) were set up by ISTC to improve coordination between adjacent projects and to encourage international collaboration. The European Commission was the first to use this. The CEG members - experts from the national institutes and industry - evaluated and managed the projects' scientific results from initial stage of proposal formulation until the final reporting. They were often involved directly in the project's details by joining the Steering Committees of the project. The Contact Expert Group for Severe Accidents and Management (CEG-SAM) is one of these groups, five project groups from this area from the total of 30 funded projects during 10 years of activity are detailed to demonstrate this: (1) QUENCH-VVER from RIAR, Dimitrovgrad and IBRAE, Moscow, and PARAMETER projects (SF1-SF4) from LUCH, Podolsk and IBRAE, Moscow; these concerned a detailed study of bundle quenching from high temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems.

  • 20.
    Breijder, Paul
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of Advanced Fuel Behaviour during Loss of Coolant Accident in Swedish Boiling Water Reactor2011Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    In accident analysis regarding nuclear power plants, it is very common to use thermal hydraulic system codes, such as TRACE, developed by U.S. NRC. In the case of licensing a power plant, this is one of the necessities.

    TRACE is a relatively new thermal hydraulic system code and a lot of knowledge is needed to implement it in a correct way, especially in accident analysis, where it is a requirement that the rules and statements in Appendix-K, dealing with criteria for ECCS-models, are modelled. In this thesis an improved model of a Swedish Boiling Water Reactor within TRACE is realized and tested. Afterwards, once a working and representative model has been obtained, a sensitivity study in conducted in order to investigate the sensitivity of TRACE for a couple of thermal hydraulic parameters.

    The sensitivity study is focussing on the eect of the peak cladding temperature, as well as the coolability of the nuclear fuel in terms of quenching and quench-front velocities.

    It is found to be hard to say unilaterally what the eect of changing a certain number of parameters on the reactor behaviour is. As it turns out to be, although strongly related, the peak cladding temperatures and the quench phenomena can behave dierently

  • 21. 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.

  • 22. 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.

  • 23.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    RELAP5 performance in predicting critical power in a BWR fuel bundle2006In: Transactions of the American Nuclear Society, 2006, p. 650-651Conference paper (Refereed)
  • 24.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Relating system-to-CFD coupled code analyses to theoretical framework of a multiscale method2008In: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work", 2008, p. 2959-2967Conference paper (Refereed)
    Abstract [en]

    Over past decades, analyses of transient processes and accidents in a nuclear power plan t have been performed, to a significant extent and with an admirable success, by means of so called system codes, e.g. RELAP5, CATHARE, ATHLET codes. These computer codes, based on a multi-fluid model of two-phase flow, provide an effective, one-dimensional description of the coolant thermal-hydraulics in the reactor system. For some components in the system, wherever needed, the effect of multi-dimensional flow is accounted for through approximate models. The later are derived from scaled experiments conducted for selected accident scenarios. Increasingly, however, we have to deal with newer and ever more complex accident scenarios. In some such cases the system codes fail to serve as simulation vehicle, largely due to its deficient treatment of multi-dimensional flow (in e.g. downcomer, lower plenum). Enter Computational Fluid Dynamics (CFD). Based on solving Navier-Stokes equations, CFD codes have been developed and used, broadly, to perform analysis of multi-dimensional flow, dominantly in non-nuclear industry and for single-phase flow applications. Although not always straightforward, CFD codes can be, and have been, used to analyze thermo-fluid processes in a certain component of the reactor system at a well-defined point during the accident progression. It is natural to think that CFD codes provide the much-needed complementary capability to the system codes. Furthermore, due to the CFD excessive demand on computational resources, ideas were proposed, and attempts were reported in the literature, to use a coupled system-to-CFD code to maximize the benefit of both tools. Easy as it might sound, progress in this area has been sluggish. In this paper, we take a close look at the progress in coupled system-to-CFD code analyses, including coupling algorithms, their implementation and performance. Tackling thermo-fluid dynamics at largely different scales, system codes and CFD codes employ different models and governing equations. This notion led us to the idea to examine the system-to-CFD coupling in the language of multiscale simulations. As a theoretical framework, we bring to bear the heterogeneous multiscale method pioneered by E and Engquist and problem classification offered by E et al.[16]. Viewing system-to-CFD coupling as a multiscale problem, the ultimate objective of the present effort is to define requirements for models and numerical methods, and develop suggestions on a coupling strategy that ensures robust and effective generation and transfer of information between scale-specific simulations (system and CFD).

  • 25.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A closure-on-demand approach to the coupling of CFD and system thermal-hydraulic codes2008In: The 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7), 2008Conference paper (Refereed)
  • 26.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Study of algorithmic requirements for a system-to-CFD coupling strategy2008In: Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS), 2008Conference paper (Refereed)
    Abstract [en]

    Over the last decades, the analysis of transients and accidents in nuclear power plants has beenperformed by system codes. Though they will remain the analyst’s tool of choice for the foreseeablefuture, their limitations are also well known. It has been suggested that an improvement in thesimulation technology can be obtained by “coupling” system codes with Computational FluidDynamics (CFD) calculations. This is usually attempted in a domain decomposition fashion: the CFDsimulation is only performed in a selected subdomain and its solution is “matched” with the systemcode solution at the interface. However, another coupling strategy can be envisioned. Namely, CFDsimulations can be used to provide closures to a system code.This strategy is based on the following two assumptions. The first assumption is that there aretransients which cannot be simulated by system codes because of the lack of adequate closures. Thesecond assumption is that appropriate closures can be provided by CFD simulations. In this paper,such a coupling strategy, inspired by the Heterogeneous Multiscale Method (HMM), is presented. Thephilosophy underlying this strategy is discussed with the help of a computational example.

  • 27.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of a “Coupling-by-Closure” approach between CFD and System Thermal-Hydraulics Codes2009In: Proc. The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), 2009Conference paper (Refereed)
  • 28.
    Cadinu, Francesco
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of In-Vessel Coolability and Retention with Control Rod Guide Tube Cooling in Boiling Water Reactors2009Conference paper (Refereed)
  • 29. Carlson, A.
    et al.
    Lakehal, D.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A Multiscale Approach for Thin-Film Slug Flow2009In: Proceedings of the 7th World Conference on Experimental Heat Transfer, Fluid Mechanics and Thermodynamics, ExHFT-7, 2009Conference paper (Refereed)
    Abstract [en]

    A multiscale modeling approach is presented for multiphase flow phenomena featuring a thin-film bounding two phases. A Micro Scale Solver predicts the thin film dynamics, influenced by an antagonistic Van der Waals force and a stabilizing repulsive force, which is mapped onto a Macro Scale Solver through a multiscale coupling. Numerical experiments of thin-film slug flow in a micropipe demonstrate that the key to capture multiscale phenomena lies in the accurate modelling of the microscale parameters. Faitful results are obtained with the multiscale treatment for the modelling of slug flow with a 10.4 nm thin-film, where pure computational multi-fluid dynamics is deficient. 

  • 30.
    Carlson, Andreas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Narayanan, C.
    ASCOMP GmbH, Technoparkstrasse 1, 8005 Z¨urich, Switzerland.
    Prediction of Two-Phase Flow in Small Tubes: A Systematic Comparison of State-of-The-Art CMFD Codes2008In: 5th European Thermal-Sciences Conference (EUROTHERM), 2008Conference paper (Refereed)
    Abstract [en]

    Multiphase dynamics and its characteristics for two-phase gas-liquid flow have been investigatedby means of advanced numerical simulations. Although important in many engineering applications, methods for robust and accurate simulations for high density and viscosity ratios remainelusive. A comprehensive comparison of two state-of-the-art Computational Multi–Fluid Dynamics (CMFD) codes, Fluent and TransAT, have been performed. The two commonly usedmethods for two–phase flow simulations, namely Volume of Fluid implemented in Fluent andLevel Set implemented in TransAT, could be compared as a result. Significant differences wereobserved between the two flow topologies predicted by the two codes. For the bubbly flow case,a recirculating flow was predicted inside the bubbles by TransAT, meanwhile no significantrecirculation was observed in the solution with Fluent. For the slug flow case a significantdeviation was observed between the results from Fluent and TransAT on the slug formationand frequency. Periodic slug formation was observed with TransAT, in agreement with theexperimental result of Chen et al. [4]. A periodic slug formation was not obtained with Fluent.

  • 31.
    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.

  • 32. 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.

  • 33. 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.

  • 34.
    Concilio Hansson, Roberta
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Experimental Study on the Dynamics of a Single Droplet Vapor Explosion2010Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    The present study aims to develop a mechanistic understanding of the thermal-hydraulic processes in a vapor explosion, which may occur in nuclear power plants during a hypothetical severe accident involving interactions of high-temperature corium melt and volatile coolant. Over the past several decades, a large body of literature has been accumulated on vapor explosion phenomenology and methods for assessment of the related risk. Vapor explosion is driven by a rapid fragmentation of high temperaturemelt droplets, leading to a substantial increase of heattransfer areas and subsequent explosive evaporation of the volatile coolant. Constrained by the liquid-phase coolant, the rapid vapor production in the interaction zone causes pressurization and dynamic loading on surrounding structures. While such a general understanding has been established, the triggering mechanism and subsequent dynamic fine fragmentation have yet not been clearly understood. A few mechanistic fragmentation models have been proposed, however, computational efforts to simulate the phenomena generated a large scatter of results.

    Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) are investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography, called SHARP (Simultaneous High-speed Acquisition of X-ray Radiography and Photography). After an elaborate image processing, the SHARP images depict the evolution of both melt material (dispersal) and coolant (bubble dynamics), and their microscale interactions, i.e. the triggering phenomenology.

    The images point to coolant entrainment into the droplet surface as the mechanism for direct contact/mixing ultimately responsible for energetic interactions. Most importantly, the MISTEE data reveals an inverse correlation between the coolant temperature and the molten droplet deformation/prefragmentation during the first bubble dynamics cycle. The SHARP observations followed by further analysis leads to a hypothesis about a novel phenomenon called pre-conditioning, according to which dynamics of the first bubble-dynamics cycle and the ability of the melt drop to deform/pre-fragment dictate the subsequent explosivity of the so-triggered droplet.

    The effect of non-condensable gases on the perceived mechanisms was investigated on the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning and interaction energetics. Analysis showed twomain aspects when compared to the MISTEE test series (withoutentrapped air). First, the investigation showed that the meltpreconditioning still strongly depends on the coolant subcooling. Second,in respect to the energetics, the tests consistently showed a reducedconversion ratio compared to that of the MISTEE test series.

    The effect of the melt material in the steam explosion triggerability was also summoned, since it would in principle directly implicate the melt preconditioning. Since a number of the thermo-physical properties of the material would influence the triggering process, we focused on the material properties by using the same dioxide material with difference concentrations, i.e. eutectic and non-eutectic. Unfortunately, due to the high melt superheat the possible differences were not perceived. Thus, inaddition to other materials, lower melt superheat tests were schedule inthe future.

  • 35.
    Concilio Hansson, Roberta
    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.
    Manickam, Louis
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A study of the effect of binary oxide materials in a single droplet vapor explosion2013In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 264, p. 168-175Article in journal (Refereed)
    Abstract [en]

    In an effort to explore fundamental mechanisms that may govern the effect of melt material on vapor explosion's triggering, fine fragmentation and energetics, a series of experiments using a binary-oxide mixture with eutectic and non-eutectic compositions were performed. Interactions of a hot liquid (WO3-CaO) droplet and a volatile liquid (water) were investigated in well-controlled, externally triggered, single-droplet experiments conducted in the Micro-interactions in steam explosion experiments (MISTEE) facility. The tests were visualized by means of a synchronized digital cinematography and continuous X-ray radiography system, called simultaneous high-speed acquisition of X-ray radiography and photography (SHARP). The acquired images followed by further analysis indicate milder interactions for the droplet with non-eutectic melt composition in the tests with low melt superheat, whereas no evident differences between eutectic and non-eutectic melt compositions regarding bubble dynamics, energetics and melt preconditioning was observed in the tests with higher melt superheat.

  • 36. Curnier, F.
    et al.
    Marquès, M.
    Kumar, Ranjan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bama, Z.
    Rychkov, V.
    Symbiosis of static and dynamic probabilistic approaches to support the design process and evaluate the safety of a SFR2015In: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015, American Nuclear Society, 2015, p. 448-453Conference paper (Refereed)
    Abstract [en]

    ASTRID, the Advanced Sodium Technological Reactor for Industrial Demonstration, is a GEN IV technological demonstrator to be commissioned near the end of the 2020 decade. The aim is to demonstrate the progress made in the field of Sodium Fast Reactor technology on an industrial scale, by qualifying innovative options, especially those pertaining to safety and operability. An original combined methodology for probabilistic safety assessment (PSA) is being developed by the CEA and its partners, AREVA NP and EDF at the conceptual design stage of ASTRID. It consists at first, of a static level 1 PSA based on the conventional fault trees (FT)/event trees (ET) approach, taking into account a time period of a week without repair of component malfunctions. Its goal is to provide probabilistic insights in the assessment of design choices and to suppress the weaknesses of the design in terms of safety considerations. A reference configuration of the safety systems is evaluated in order to identify dominant accident sequences. Sensitivity studies are then performed on various design alternatives to define the optimal safety systems configurations that will minimize core damage frequency. It takes into account recent design evolutions for decay heat removal (DHR) systems and support systems, and re-evaluates the preliminary results from ASTRID PSA modeling. The conventional FT/ET approach initially developed for PWRs (Wash 1400) appears to be unsuitable for Sodium Fast Reactors (SFR) PSA because: This approach is binary and static, The probabilistic study for SFR cannot be limited to short periods of time - when repair is not possible - because several months are necessary for the thermal leakage to be equivalent to decay heat, SFR technology cannot rely simply on DHR complementary systems, The modeling by FT/ET is not designed for long periods of time, Repair, on along and middle term basis, of failed components is not considered. Therefore, dynamic PSA approaches have been investigated to extend the conventional PSA to longer periods of time by taking into account the specific characteristics of a sodium reactor such as its great thermal inertia - which allows the operator to make interventions - and the fact that sodium circuits present risks of irreversible and temperature-sensitive failures. What these approaches have in common is the possibility of taking into account the repair of failed components. Simplified thermal-hydraulic calculations were performed to characterize the reactor at any given moment in the accident scenario. The benefits of dynamic approaches on short periods of time will be quantitatively evaluated in 2015.

  • 37. Demazière, C.
    et al.
    Dykin, V.
    Hernández-Solís, Augusto
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Boman, V.
    Modelling of stationary fluctuations in nuclear reactor cores in the frequency domain2015In: Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015, American Nuclear Society, 2015, p. 2406-2419Conference paper (Refereed)
    Abstract [en]

    This paper presents the development of a numerical tool to simulate the effect of stationary fluctuations in Light Water Reactor cores. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool calculates the three-dimensional space-frequency distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the homogeneous equilibrium model complemented with pre-computed static slip ratio. Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool is currently entirely Matlab based with input data provided by an external static core simulator. The paper also presents the results of dynamic simulations performed for a typical pressurized water reactor and for a typical boiling water reactor, as illustrations of the capabilities of the tool.

  • 38. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum2015In: International Conference on Nuclear Engineering, Proceedings, ICONE, JSME , 2015Conference paper (Refereed)
    Abstract [en]

    MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.

  • 39. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Extension of the MELCOR code for analysis of late in-vessel phase of a severe accident2015In: IYCE 2015 - Proceedings: 2015 5th International Youth Conference on Energy, IEEE conference proceedings, 2015Conference paper (Refereed)
    Abstract [en]

    The simulation of severe accidents in nuclear power plants with system codes is a powerful tool to improve the safety measures to prevent severe accidents. The further development of severe accident codes is part of current research. MELCOR, as the leading nuclear safety code, provides the possibility to be coupled to other codes. A detailed knowledge of this coupling interface is necessary to use this possibility. Therefore, the software tool DINAMO, which contains the coupling routines and an interface to communicate with other programs, was developed. Using DINAMO it is possible to utilize new models for specific phenomena in MELCOR. In the present work the Phase-Change Effective Convectivity Model was coupled using the CFD-software OpenFOAM and DINAMO to MELCOR to improve the prediction of molten core material in the lower plenum of a reactor pressure vessel. The simulation results were compared to the experimental findings of the LIVE-facility.

  • 40.
    Dinh, Truc Nam
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Nourgaliev, R. R.
    Theofanous, T. G.
    On the numerical simulation of acceleration-driven multi-fluid mixing2006In: Multiphase Science and Technology, ISSN 0276-1459, E-ISSN 1943-6181, Vol. 18, no 3, p. 199-230Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with computational prediction of acceleration-induced multi-fluid mixing phenomena. Premises and performance of existing approaches are reviewed and analyzed with focus on a late phase behavior. We introduce a new framework whose central idea is to use an interfacial area transport equation (IATE) and a subgrid scale model (SGS) of multi-fluid turbulence to provide a natural transition from DNS-based simulation toward an effective-field model (EFM) and deeply into well-mixed states with continuous refinement of length scale. We present new results and important insights derived from our work on four platform technologies: DNS, EFM, IATE and SGS. We discuss the approach to ensure that developments in different areas effectively emerge and function seamlessly in an overall computational platform for multi-fluid mixing.

  • 41.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Material property effect in steam explosion energetics: Revisited2007In: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12, 2007Conference paper (Refereed)
    Abstract [en]

    Steam explosion, as a threat to LWR reactor vessel and containment integrity, has been postulated to occur during a hypothetical severe accident with relocation of molten core materials to a water pool either in-vessel or ex-vessel. Studies of molten fuel-coolant interactions (FCI) conducted over the past decades have not resolved the controversy about whether, when, and how melt material properties influence steam explosion energetics. Crucial questions persist about safety significance of experimental evidence about corium low explosivity in various reactor accident scenarios. In this paper, taking into consideration results from recent FCI experiments and analyses, we revisit the study of Dinh et al (1998) and hypotheses proposed therein about mechanisms by which corium physical properties may influence steam explosions. Corium high density, high melting point and low conductivity are found to be central to mechanisms in premixing that govern corium low explosivity. For micro-interactions, three processes, namely drop surface undercooling, nucleation and growth of solid phases, and interfacial instability and breakup are evaluated with respect to their role in fine fragmentation. The paper provides a new hypothesis for rationalizing the effect of corium composition (eutectic vs. non-eutectic) on its triggerability and energetics.

  • 42.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Multiphase flow phenomena of steam generator tube rupture in a lead-cooled reactor system: A scoping analysis2008In: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work", 2008, p. 2765-2775Conference paper (Refereed)
    Abstract [en]

    The paper is concerned with understanding and quantification of intense multiphase interactions in a Steam Generator Tube Rupture (SGTR) scenario in advanced lead-cooled reactor systems. The scoping approach taken in this paper is to focus on key flow physics that complements other ongoing detailed computational and experimental efforts on SGTG analysis. The present study suggests that (i) the initial pressure shock wave poses no credible threat to invessel structures, except for limited pressure loading on very few adjacent heat-exchange tubes; (ii) the sloshing-relatedfluid motion is well bounded in a domain beyond the heat exchanger; (iii) the pre-mixture is not pre-conditioned for triggering and a postulated steam explosion would have limited energetics; and (iv) an initial discharge of steam/water mass is amenable for entrapment in the primary coolant flow. Implications for further research are discussed in the paper.

  • 43.
    Dinh, Truc-Nam
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Department of Nuclear Science and Engineering, Idaho National Laboratory, United States .
    Hansson Concilio, Roberta
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On solidification mechanism that governs the effect of binary melt composition on steam explosion energetics2008In: Transactions of the American Nuclear Society, 2008, p. 615-616Conference paper (Refereed)
  • 44.
    Dinh, Truc-Nam
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Hansson, Roberta
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On Solidification Mechanism that Govern the Effect of Binary Melt Composition on Steam Explosion Energetics2008In: Transaction of American Nuclear Society 2008, American Nuclear Society, 2008, p. 615-616Conference paper (Refereed)
  • 45. Dokhane, A.
    et al.
    Judd, J.
    Gajev, Ivan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zerkak, O.
    Ferroukhi, H.
    Kozlowski, T.
    Analysis of Oskarshamn-2 stability event using TRACE/SIMULATE-3K and comparison to TRACE/PARCS and SIMULATE-3K stand-alone2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 102, p. 190-199Article in journal (Refereed)
    Abstract [en]

    With the goal to enhance the capability to perform best-estimate simulations of Light Water Reactors (LWRs) transients, with strong coupling between core neutronics and plant thermal-hydraulic, a coupling between TRACE and SIMULATE-3K (TS3K) was developed in collaboration between PSI and Studsvik for analyses involving interactions between system and core. In order to verify the coupling scheme and the coupled code capabilities to simulate complex transients, the OECD/NEA Oskarshmn-2 (O-2) Stability benchmark was modeled with the coupled code TS3K. The main goal of this paper is to present TS3K analyses of the Oskarshamn-2 stability event, noting that this constitutes the first reported assessment of this code system for a BWR stability problem. A systematic analysis is carried out using different time-space discretization schemes in order to identify an optimized methodology to simulate correctly the O-2 stability event. In this context, the TS3K results are compared to the available benchmark data both for steady-state and transient conditions. The results show that using a refined model in space and time, the TS3K model can successfully capture the entire behavior of the transient qualitatively, i.e. onset of the instability with growing oscillation amplitudes, as well as quantitatively, i.e. Decay Ratio and resonance frequency. In addition, the results are compared also to those obtained using TRACE/PARCS and S3K stand-alone, which allows a systematic comparison between different codes.

  • 46.
    Dombrovsky, L.A.
    et al.
    Joint Institute for High Temperatures, Moscow, Russia.
    Davydov, M.V.
    Electrogorsk Research & Engineering Center on NPP Safety, Saint Constantine 6,142530, Electrogorsk, Moscow region, Russia.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal radiation modeling in numerical simulation of melt-coolant interaction2008In: Proc. Int. Symp. Adv. Comput. Heat Transfer (CHT-08), 2008Conference paper (Refereed)
    Abstract [en]

    This paper is concerned with radiation heat transfer modeling in multiphase disperse systems, which are formed in high-temperature melt-coolant interactions. This problem is important for complex interaction of the core melt with water in the case of a hypothetical severe accident in light-water nuclear reactors. The nonlocal effects of thermal radiation due to the semitransparency of water in the visible and near-infrared spectral ranges are taken into account by use of the recently developed large-cell radiation model (LCRM) based on the spectral radiation energy balance for single computational cells. In contrast to the local approach for radiative heating of water by particles (OMM—opaque medium model), the LCRM includes radiative heat transfer between the particles of different temperatures. The regular integrated code VAPEX-P, intended to model the premixing stage of FCI, was employed for verification of the LCRM in a realistic range of the problem parameters. A comparison with the OMM and the more accurate P1 approximation showed that the LCRM can be recommended for the engineering problem under consideration. The effects of the temperature difference in solidifying particles are analyzed by use of the recently suggested approximation of transient temperature profile in the particles. It is shown that the effect of the temperature difference on heat transfer from corium particles to ambient water is considerable and should not be ignored in the calculations. An advanced computational model based on the LCRM for the radiation source function and subsequent integration of radiative transfer equation along the rays is also discussed.

  • 47.
    Dombrovsky, L.A.
    et al.
    Joint Institute for High Temperatures, Krasnokazarmennaya 17A, 111116, Moscow, Russian Federation.
    Davydov, M.V.
    Electrogorsk Research and Engineering Center on NPP Safety, Saint Constantine 6, 142530, Electrogorsk, Moscow region, Russian Federation.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal radiation modeling in numerical simulation of melt-coolant interaction2009In: Computational Thermal Sciences, ISSN 1940-2503, Vol. 1, no 1, p. 1-35Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with radiation heat transfer modeling in multiphase disperse systems, which are formed in high-temperaturemelt-coolant interactions. This problem is important for complex interaction of the core melt with water in the case of a hypothetical severe accident in light-water nuclear reactors. The nonlocal effects of thermal radiation due to the semitransparency of water in the visible and near-infrared spectral ranges are taken into account by use of the recently developed large-cell radiation model (LCRM) based on the spectralradiation energy balance for single computational cells. In contrast to the local approach for radiative heating of water by particles (OMMopaque medium model), the LCRM includes radiative heat transfer between the particles of different temperatures. The regular integrated code VAPEX-P, intended to model the premixing stage of FCI, was employed for verification of the LCRM in a realistic range of the problem parameters. A comparison with the OMM and the more accurate P1 approximation showed that the LCRM can be recommended for the engineering problem under consideration. The effects of the temperature difference in solidifying particles are analyzed by use of the recently suggested approximation of transient temperature profile in the particles. It is shown that the effect of the temperature difference on heat transfer from corium particles to ambient water is considerable and should not be ignored in the calculations. An advanced computational model based on the LCRM for theradiation source function and subsequent integration of radiative transfer equation along the rays is also discussed. 

  • 48. Dombrovsky, Leonid A.
    et al.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The effect of thermal radiation on the solidification dynamics of metal oxide melt droplets2008In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 238, no 6, p. 1421-1429Article in journal (Refereed)
    Abstract [en]

    Cooling and solidification of metal oxide droplets in water are considered, using a single-particle model which takes into account heat conduction and thermal radiation transfer within the particle. It is shown that, for millimeter-size particles, near-infrared absorption of the particle's substance determines the solidification pattern and dynamics. For semi-transparent aluminum oxide particles, the rate of surface solidification is controlled by convective heat transfer. For opaque corium particles, thermal radiation from the particle surface leads to fast surface solidification. The impact of so-formed crust layer on subsequent particle fragmentation is discussed with respect to its influence on steam explosion.

  • 49. Fichot, F.
    et al.
    Carénini, L.
    Sangiorgi, M.
    Hermsmeyer, S.
    Miassoedov, A.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zdarek, J.
    Guenadou, D.
    Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, p. 36-45Article in journal (Refereed)
    Abstract [en]

    The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

  • 50. Fischer, M.
    et al.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bezlepkin, V. V.
    Hamazaki, R.
    Miassoedov, A.
    Core melt stabilization concepts for existing and future LWRs and associated R&D needs2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, Vol. 9, p. 7578-7592Conference paper (Refereed)
    Abstract [en]

    In the event of a severe accident with core melting in a NPP the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage of internal structures. The related failure modes may result in significant long-term radiological consequences and high related costs. Because of this, the licensing framework of several countries now includes the request to implement mitigative core melt stabilization measures. This does not only apply to new builds but also to existing LWR plants. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles like: melt fragmentation in a deep water pool or during molten core concrete interaction with top-flooding, water injection from the bottom (COMET concept), and retention in an outside-cooled crucible structure. The provided overview covers the physical background and functional principles of these concepts, as well as their status of validation and, if applicable, the remaining open issues and R&D needs. For concepts based on melt retention inside a cooled crucible that reached sufficient maturity to be implemented in current Gen-III+ designs, like the VVER-1000/1200 and the EPR™, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.

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