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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.

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  • 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.
    Bandaru, S V Ravikumar
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Komlev, Andrei A.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sköld, Per
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Upward-facing multi-nozzle spray cooling experiments for external cooling of reactor pressure vessels2020In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 163, article id 120516Article in journal (Refereed)
    Abstract [en]

    Cooling by water spray is a well-known technology that can reach significantly higher Critical Heat Flux (CHF) compared to other cooling methods. For the light water reactor safety, the in-vessel retention (IVR) by external reactor vessel cooling (ERVC) is a comprehensive severe accident management strategy to arrest and confine the corium in the lower head of the reactor pressure vessel. Heat fluxes up to 1.5 MW/m2 have already been assumed attainable in low-power nuclear reactors while cooling required in high-power reactors is expected to reach 2.5 MW/m2. Instead of reactor lower head flooding and relying on cooling due to natural convection, a viable and more efficient alternative is to spray the external surface of the vessel. Given all the advantages of spray cooling reported in the literature, a lab-scale experimental facility was built to validate the efficiency of multi-nozzle spray cooling of a downward-facing heated surface inclined at different angles up to 90o. The facility employed a 2×3 matrix of spray nozzles to cool the FeCrAl alloy foil with an effectively heated surface area of 96 cm2 using water as the coolant. Heat loads and surface inclinations were varied parameters in the test matrix. The results show that no significant variations in spray cooling performance concerning the inclination of the heated surface. A surface heat flux of 2.5 MW/m2 was achieved at every inclination of the downward-facing surface. The results also indicate that more uniform liquid film distribution could be obtained for some inclinations, which in turn leads to maintaining low surface temperature. The obtained surface heat flux margin by spray cooling indicates that it is feasible to adopt IVR-ERVC strategy for a large power reactor.

  • 7.
    Bandaru, S V Ravikumar
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thakre, Sachin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Multi-nozzle spray cooling of a reactor pressure vessel steel plate for the application of ex-vessel cooling2021In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 375, article id 111101Article in journal (Refereed)
    Abstract [en]

    Spray cooling is a versatile technology for various cooling applications involving high surface heat fluxes. Experimental facility was built to study heat transfer performance of an upward multi-nozzle array of water sprays impacting a surface of heated plate made of reactor vessel grade steel. The effect of inclination angles of the steel surface on the cooling performance was investigated to assess heat transfer in complex semispherical/ semielliptical geometry of large reactor lower head and to address possible application of spray cooling in severe accident management (SAM) of light water reactors (LWRs) based on In-vessel melt retention with external reactor vessel cooling (IVR-ERVC). Joule heating of SA302B steel foil of 0.15 mm thickness and surface area of 96 cm2 enabled prototypic heat fluxes to be evacuated during reactor accident. A 2×3 array of full jet narrow-coned pressure-swirl spray nozzles was used to reproduce multi-nozzle cooling. The tests were conducted as a series of consequent steady states realized at stepwise increasing power and surface heat fluxes up to the maximum values of 29 kW and 2.97 MW/m2 limited in the specific facility design. Seven surface inclinations, between 0o and 90o were tested and no significant variations in spray cooling performance with the inclination of the heated surface was found. The results indicated a promising prospect of using a multi-nozzle array system for cooling of large surface area of reactor lower head. Much higher heat fluxes can be safely extracted by spray cooling in comparison with the critical heat fluxes that appeared at RPV water pool cooling at natural convection. The maximum value of heat flux at direct spray impact zones and its drop-off slightly from the center to the periphery of the spray cone was detected in the tests. The water flow rate and liquid subcooling significantly influenced maximum steel surface temperature but had no noticeable effects on surface temperature uniformity. The spray-to-spray interaction had no observable effects on local surface temperatures, however, the colliding zones where four spray cones have visible effects on local surface temperatures due to poor liquid momentum. The results also showed that more uniform liquid film distribution could be obtained for some inclinations because of improved liquid drainage, which in turn leads to maintaining low surface temperatures. 

  • 8. 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. 

  • 9. 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.

  • 10.
    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.

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  • 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.
    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.

  • 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.
    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.

  • 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.
    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.

  • 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.
    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.

  • 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.
    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.

  • 16.
    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.

  • 17.
    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
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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.

  • 18.
    Bechta, Sevostian
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Miassoedov, Alexei
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Journeau, Christophe
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Okamoto, Koji
    Univ Tokyo, Tokyo, Japan..
    Manara, Dario
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Bottomley, David
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Kurata, Masaki
    JAEA CLADS Lab, Iwaki, Fukushima, Japan..
    Sehgal, Bal Raj
    Stuckert, Jun
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Steinbrueck, Martin
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Fluhrer, Beatrix
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Keim, Torsten
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Fischer, Manfred
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Langrock, Gert
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Piluso, Pascal
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Hozer, Zoltan
    MTA EK, Budapest, Hungary..
    Kiselova, Monika
    UJV REZ As, Hlavni 130, F-25068 Husinec Rez, Czech Republic..
    Belloni, Francesco
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    Schyns, Marc
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    On the EU-Japan roadmap for experimental research on corium behavior2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed)
    Abstract [en]

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

  • 19.
    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.

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  • 20.
    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.

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  • 21.
    Bian, Boshen
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    CFD Study of Molten Pool Convection in a Reactor Vessel during a Severe Accident2023Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    During severe accidents in nuclear reactors, the core and internal structures can melt down and relocate into the reactor pressure vessel (RPV) lower head (LH) forming there a stratified molten corium pool. The pool generally consists of superheated oxidic and metallic liquid layers imposing thermo-mechanical loads on the RPV wall. The in-vessel retention (IVR) strategy employs external cooling with water to maintain RPV integrity. Investigating the thermo-fluid behaviour of corium and predicting heat flux distribution on the vessel wall are crucial. The molten pool exhibits natural convection, which can typically consist of two stratified layers. There exists internally heated (IH) natural convection in the oxidic layer and Rayleigh-Bénard (RB) convection in the surface metallic layer.

    This study starts by illustrating the mathematical models that involve the numerical study of natural convection flow in molten corium. A verification work of the model has been done using a previous direct numerical simulation (DNS) study, and the results show good agreement. In addition, a scaling theory of the natural convection flow is demonstrated to facilitate the pre-estimation based on the Rayleigh number (Ra) and Prandtl number (Pr). After that, the numerical approaches involved in the numerical simulation of the corium are illustrated, especially focusing on the DNS method. A DNS mesh strategy is proposed in the form of a pipeline from the pre-estimation to the post-check. A scalability study of Nek5000 is performed on four different HPC clusters based on a DNS case of the IH molten convection in a hemispherical geometry with Ra=1.6×1011. The results show a super-liner speedup property of Nek5000 on each cluster within a certain range.

    Then, three numerical studies focusing on turbulent natural convection flow within both the oxidic and metallic layers of corium are demonstrated and discussed. Through these simulations, the thermos-fluid behaviour of the system is examined in detail, including flow configuration, temperature distribution, heat flux profiles on cooling boundaries, and turbulent quantities.

    1. A DNS investigation is performed on the IH molten pool convection within a hemispherical domain, employing a Rayleigh number of 1.6×1011 and a Prandtl number of 0.5. The results show a turbulent flow characterized by three distinct regions, consistent with the observation from the BALI experiments. Detailed information regarding turbulence, including turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance, is presented. Furthermore, the study offers comprehensive 3D heat flux distributions along the boundaries, showing heat flux fluctuations along the top boundary due to nearby turbulent eddies and a nonlinear increase in heat flux along the curved boundary from bottom to top.

    2. A numerical study investigates the effect of Prandtl number on the natural convection of an IH molten pool in a 3D semi-circular test section. Prandtl numbers of 3.11, 1.0, and 0.5 are considered, with a Ra= 6.54×1011. Smaller Prandtl numbers result in more vigorous turbulent motion and a thicker layer of intense turbulent mixing in the upper region. The descending flow extends further down the bottom, creating a stronger circulation at the bottom with smaller Pr. Additionally, smaller Pr leads to more thermal stripping structures and less stable stratification layers. Comparing heat fluxes on the top and curved walls reveals higher fluctuation frequency with smaller Pr for heat fluxes to the top boundary. However, the maximum heat fluxes to the side walls are lower with smaller Pr.

    3. A numerical study investigates the turbulent natural convection in a 3D fluid layer based on the BALI-Metal 8U experiment. Different methods, including DNS and three Reynolds-averaged Navier-Stokes (RANS) models, are employed. The results are compared with experimental data, and the performance of the RANS models is evaluated using DNS as a reference. DNS reproduces a two-distinct region flow structure observed in experiments, while the k-ω SST model exhibits similar flow patterns and TKE profiles. However, all simulations overpredict temperature compared to experimental data, with DNS providing the closest results. The DNS results also achieve better agreement with experimental data in terms of heat flux distribution and energy balance, specifically capturing the transient maximum heat flux on the lateral cooling wall. This transient behaviour plays a crucial role in accurately estimating the ‘focusing effect’.

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  • 22.
    Bian, Boshen
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Daniele, Dovizio
    Nuclear Research and Consultancy Group.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Direct numerical simulation of internally heated natural convection in a hemispherical geometryManuscript (preprint) (Other (popular science, discussion, etc.))
    Abstract [en]

    Internally heated (IH) natural convection can be found in nature, industrial processes, or during a severe accident in a light water reactor. In this accident scenario, the nuclear reactor core and some internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV) and interact with the remaining coolant. Subsequent re-heating and re-melting under decay and oxidation heat creates a transition from a debris bed to a molten pool. The molten pool, which can involve more than hundred tons of dangerously superheated oxidic and metallic liquids, imposes thermo-mechanical loads on the vessel wall that can lead to a thermal and/or structural failure of the vessel and subsequent release of radioactive materials to the reactor pit, and can possibly make its way to the environment. This study uses Direct Numerical Simulation (DNS) to investigate homogeneous IH molten pool convection in a hemispherical domain using Nek5000, an open-source spectral element code. With a Rayleigh number of 1.6×1011, the highest reached through DNS in this confined hemispherical geometry, and a Prandtl number of 0.5, which corresponds to a prototypic corium, the study provides detailed information on the thermo-fluid behaviour. The results show a turbulent flow with three distinct regions, consistent with the general flow observations from the BALI experiments. The study also presents detailed information on turbulence, such as turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance. Additionally, the study provides 3D heat flux distributions along the boundaries. The heat fluxes along the top boundary fluctuate due to the turbulent eddies in the vicinity, while along the curved boundary the heat fluxes increase nonlinearly from the bottom to the top.

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  • 23.
    Bian, Boshen
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dovizio, Daniele
    bNuclear Research and Consultancy Group, the Kingdom of the Netherlands.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Nuclear Futures Institute, Bangor University, United Kingdom.
    Direct numerical simulation of internally heated natural convection in a hemispherical geometry2024In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 220, article id 124997Article in journal (Refereed)
    Abstract [en]

    Internally heated (IH) natural convection can be found in nature, industrial processes, or during a severe accident in a light water reactor. In this accident scenario, the nuclear reactor core and some internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV) and interact with the remaining coolant. Subsequent re-heating and re-melting under decay and oxidation heat creates a transition from a debris bed to a molten pool. The molten pool, which can involve more than hundred tons of dangerously superheated oxidic and metallic liquids, imposes thermo-mechanical loads on the vessel wall that can lead to a thermal and/or structural failure of the vessel and subsequent release of radioactive materials to the reactor pit, and can possibly make its way to the environment. This study uses Direct Numerical Simulation (DNS) to investigate homogeneous IH molten pool convection in a hemispherical domain using Nek5000, an open-source spectral element code. With a Rayleigh number of 1.6 × 1011, the highest reached through DNS in this confined hemispherical geometry, and a Prandtl number of 0.5, which corresponds to a prototypic corium, the study provides detailed information on the thermo-fluid behavior. The results show a turbulent flow with three distinct regions, consistent with the general flow observations from the BALI experiments. The study also presents detailed information on turbulence, such as turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance. Additionally, the study provides 3D heat flux distributions along the boundaries. The heat fluxes along the top boundary fluctuate due to the turbulent eddies in the vicinity, while along the curved boundary the heat fluxes increase nonlinearly from the bottom to the top.

  • 24.
    Bian, Boshen
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gong, J.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Scalability of Nek5000 on High-Performance Computing Clusters Toward Direct Numerical Simulation of Molten Pool Convection2022In: Frontiers in Energy Research, E-ISSN 2296-598X, Vol. 10, article id 864821Article in journal (Refereed)
    Abstract [en]

    In a postulated severe accident, a molten pool with decay heat can form in the lower head of a reactor pressure vessel, threatening the vessel’s structural integrity. Natural convection in molten pools with extremely high Rayleigh (Ra) number is not yet fully understood as accurate simulation of the intense turbulence remains an outstanding challenge. Various models have been implemented in many studies, such as RANS (Reynolds-averaged Navier–Stokes), LES (large-eddy simulation), and DNS (direct numerical simulation). DNS can provide the most accurate results but at the expense of large computational resources. As the significant development of the HPC (high-performance computing) technology emerges, DNS becomes a more feasible method in molten pool simulations. Nek5000 is an open-source code for the simulation of incompressible flows, which is based on a high-order SEM (spectral element method) discretization strategy. Nek5000 has been performed on many supercomputing clusters, and the parallel performance of benchmarks can be useful for the estimation of computation budgets. In this work, we conducted scalability tests of Nek5000 on four different HPC clusters, namely, JUWELS (Atos Bullsquana X1000), Hawk (HPE Apollo 9000), ARCHER2 (HPE Cray EX), and Beskow (Cray XC40). The reference case is a DNS of molten pool convection in a hemispherical configuration with Ra = 1011, where the computational domain consisted of 391 million grid points. The objectives are (i) to determine if there is strong scalability of Nek5000 for the specific problem on the currently available systems and (ii) to explore the feasibility of obtaining DNS data for much higher Ra. We found super-linear speed-up up to 65536 MPI-rank on Hawk and ARCHER2 systems and around 8000 MPI-rank on JUWELS and Beskow systems. We achieved the best performance with the Hawk system with reasonably good results up to 131072 MPI-rank, which is attributed to the hypercube technique on its interconnection. Given the current HPC technology, it is feasible to obtain DNS data for Ra = 1012, but for cases higher than this, significant improvement in hardware and software HPC technology is necessary.

  • 25.
    Bian, Boshen
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of DNS and RANS Approaches on Turbulent Natural Convection against the BALI-Metal ExperimentManuscript (preprint) (Other academic)
    Abstract [en]

    During severe accident scenarios in nuclear reactors, the core and internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV), where they interact with any remaining coolant. This process can lead to the formation of a stratified molten pool, which is also called corium. It consists of dangerously superheated oxidic and metallic liquids, which imposes thermo-mechanical loads on the vessel wall. Typically, the molten pool separates into distinct layers, with a lighter layer of metallic materials on top and a denser layer of oxides at the bottom. The metal layer acts as a heat sink, absorbing heat from the heat-generating oxide layer and conducting it towards the inner wall of the RPV. This concentrated heat load to the vessel is known as the focusing effect.

    This study conducts numerical simulations of the turbulent natural convection flow in a fluid layer undergoing both top and lateral cooling based on the BALI-Metal 8U experiment. Different methods were employed, including Direct Numerical Simulation (DNS) and three Reynolds-Averaged Navier-Stokes (RANS) models: k-ω SST, standard k-ε, and Reynolds stress equation model (RSM). The simulation results are compared with experimental data, and the RANS models are assessed using the DNS results. The results reveal that DNS is able to reproduce a two-distinct region flow structure similar to the experimental observations. The k-ω SST model shows similar flow patterns and turbulent kinetic energy (TKE) profile as the DNS results. Regarding the temperature field, all simulations overpredict temperature compared to the experimental data, with DNS providing the closest results. The turbulent heat flux (THF) result shows the RANS models are incapable of accurately modelling THF in turbulent natural convection flow. The heat flux analysis demonstrates that DNS achieved good agreement with experimental data in terms of heat flux distribution and energy balance, while the RANS models underestimate the focusing effect. Furthermore, DNS captures the transient maximum heat flux on the lateral cooling wall, which is higher than the time-averaged value, an important factor for estimating the focusing effect.

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  • 26.
    Bian, Boshen
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dovizio, Daniele
    Nucl Res & Consultancy Grp NRG, Arnhem, Netherlands..
    Direct numerical simulation of molten pool convection in a 3D semicircular slice at different Prandtl numbers2022In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 393, p. 111772-, article id 111772Article in journal (Refereed)
    Abstract [en]

    In this paper, a Direct Numerical Simulation (DNS) of an internally heated (IH) natural convection in a 3D semicircular slice molten pool is conducted using Nek5000, a CFD solver with spatial discretization based on the spectral element method. The mesh requirements in the bulk and boundary layers are both fulfilled using known correlations. A calculation of a simplified internally heated box is first established with an excellent agreement to existing data. Next, simulation of the 3D semi-circular is performed showing qualitative agreement with the general flow observations from the BALI experiments. The velocity field shows that the flow domain is divided into three regions, i.e., intensive turbulent eddies in the upper domain, weak flow motion in the lower domain, and the descending flow along the curved boundary. Correspondingly, the temperature field in the upper domain is relatively homogenous, while that in the lower domain is characterized by stratified layers. Further, the heat flux distribution along the boundaries shows that the heat fluxes fluctuate along the top wall due to turbulent eddies, and the heat fluxes at the curved wall increase nonlinearly from the bottom to the top. Finally, the influence of Prandtl number indicates that smaller Prandtl number will lead to more turbulence eddies, deeper descending flow, and more even redistribution of heat thereby lowering the maximum heat flux to the curved walls.

  • 27.
    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
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  • 28. 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.

  • 29.
    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

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  • 30. 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.

  • 31. 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.

  • 32.
    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)
  • 33.
    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).

  • 34.
    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)
  • 35.
    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.

  • 36.
    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)
  • 37.
    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)
  • 38. 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. 

  • 39.
    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.

  • 40.
    Chen, Yangli
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    MELCOR Capability Development for Simulation of Debris Bed Coolability2021Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    The severe accident management (SAM) strategy for a Nordic boiling water reactor (BWR) employs cavity flooding prior to vessel failure, so that the core melt (corium) discharged from the vessel could fragment and form a particulate debris bed. The key to the success of this SAM strategy is the coolability of ex-vessel debris beds.

    The safety analysis involves knowledge about the reactor response to severe accidents under this SAM strategy, which requires the integral simulation of a system code such as MELCOR. Since currently the MELCOR code lacks the modeling of ex-vessel particulate debris beds, the present study aims to develop the capability of MELCOR for the simulation of debris bed coolability through the coupling of MELCOR with other codes, which are dedicated to this phenomenon.

    The study is started from the qualification of a MELCOR model for severe accident analysis of a reference Nordic BWR, with the aim to help identify a proper core nodalization. For this purpose, three different core meshes (coarse, medium, and fine) are employed to obtain their impacts on corium release conditions. It is found the coarse mesh is sufficient in the present study, since it is not only computationally efficient, but also predicting earlier vessel failure and faster corium release, providing a more conservative condition for debris bed coolability analysis.

    Two couplings are then adopted: (i) coupling of MELCOR with the COCOMO code, which is a mechanistic code for simulation of thermal hydraulics in debris beds; and (ii) coupling of MELCOR with a surrogate model developed in the present study. The first method can simulate the transient behavior of a debris bed during quench process. The second method can efficiently predict the coolability limit (dryout power) required in safety analysis. The surrogate model is developed based on the COCOMO prediction of two-dimensional debris beds.

    The developed simulation tools, including the coupled codes and the surrogate model, are applied to the safety analysis of the reference Nordic BWR. The coupled MELCOR/COCOMO simulation is used to investigate the debris bed properties. The effective particle diameter is found as approximately 10% larger than the surface mean diameter of a debris bed with distributed sizes, quantified by the quench rate. For the effect of debris bed shape, it shows a faster quench process with a lower bed slope angle. The quench front propagation as well as the responses of local temperature and containment pressure are obtained.

    The coupled MELCOR/surrogate model simulation is performed to estimate the coolability of ex-vessel vessel debris beds. The results show that debris beds are coolable under prototypical conditions with probable bed properties. The surrogate model is used to generate coolability maps, which show the debris bed coolability with the variation of bed properties. The sensitivity analysis indicates that the porosity and the geometry are the most influential to coolability limit. An uncertainty analysis methodology is proposed to obtain the probability of non-coolable debris beds.

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  • 41.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development and application of a surrogate model for quick estimation of ex-vessel debris bed coolability2020In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 370, article id 110898Article in journal (Refereed)
    Abstract [en]

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

  • 42.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of surrogate model for debris bed coolability analysis2019In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6770-6779Conference paper (Refereed)
    Abstract [en]

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

  • 43.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

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

  • 44.
    Chen, Yangli
    et al.
    KTH.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 143, article id 107490Article in journal (Refereed)
    Abstract [en]

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

  • 45.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zhang, Huimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds2022In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 165, article id 108643Article in journal (Refereed)
    Abstract [en]

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

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  • 46.
    Chen, Yangli
    et al.
    KTH.
    Zhang, Huimin
    KTH.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed)
    Abstract [en]

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

  • 47.
    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.

  • 48. 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.

  • 49. 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.

  • 50.
    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.

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