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  • 1. 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. 

  • 2. 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.

  • 3.
    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.

  • 4.
    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.

  • 5.
    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.

  • 6.
    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.

  • 7.
    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.

  • 8.
    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.

  • 9.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Yakush, Sergey
    Institute for Problems in Mechanics, Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow, 119526, Russia.
    Kudinov, Pavel
    Validation of DECOSIM code against experiments on particle spreading by two-phase flows in water pool2016In: Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, NUTHOS-11, 2016, article id N11A0531Conference paper (Refereed)
    Abstract [en]

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

  • 10. Bulat, A.F.
    et al.
    Voloshin, A.I.
    Kudinov, Pavel
    Dnepropetrovsk State University.
    Mathematical simulation of processes of heat and mass transfer at plasma ignition of coal powder2004In: Proceedings of V Minsk International Forum on Heat and Mass Transfer. Institute of Heat and Mass Transfer A.V. Lykova, 2004Conference paper (Refereed)
  • 11.
    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)
  • 12.
    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.

  • 13.
    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)
  • 14.
    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)
  • 15. 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. 

  • 16.
    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.

  • 17.
    Dinh, Truc-Nam
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Department of Nuclear Science and Engineering, Idaho National Laboratory, United States .
    Hansson Concilio, Roberta
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On solidification mechanism that governs the effect of binary melt composition on steam explosion energetics2008In: Transactions of the American Nuclear Society, 2008, p. 615-616Conference paper (Refereed)
  • 18.
    Dinh, Truc-Nam
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Hansson, Roberta
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On Solidification Mechanism that Govern the Effect of Binary Melt Composition on Steam Explosion Energetics2008In: Transaction of American Nuclear Society 2008, American Nuclear Society, 2008, p. 615-616Conference paper (Refereed)
  • 19.
    Dombrovsky, L.A.
    et al.
    Joint Institute for High Temperatures, Moscow, Russia.
    Davydov, M.V.
    Electrogorsk Research & Engineering Center on NPP Safety, Saint Constantine 6,142530, Electrogorsk, Moscow region, Russia.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal radiation modeling in numerical simulation of melt-coolant interaction2008In: Proc. Int. Symp. Adv. Comput. Heat Transfer (CHT-08), 2008Conference paper (Refereed)
    Abstract [en]

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

  • 20.
    Dombrovsky, L.A.
    et al.
    Joint Institute for High Temperatures, Krasnokazarmennaya 17A, 111116, Moscow, Russian Federation.
    Davydov, M.V.
    Electrogorsk Research and Engineering Center on NPP Safety, Saint Constantine 6, 142530, Electrogorsk, Moscow region, Russian Federation.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal radiation modeling in numerical simulation of melt-coolant interaction2009In: Computational Thermal Sciences, ISSN 1940-2503, Vol. 1, no 1, p. 1-35Article in journal (Refereed)
    Abstract [en]

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

  • 21. Frid, W.
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ex-Vessel Melt Coolability Issue in BWRs with Deep Water Pool in Lower Drywell2010In:  , 2010Conference paper (Refereed)
  • 22.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kapulla, R
    Paranjape, S
    Paladino, D
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Modeling of Thermal Stratification and Mixing Induced by Steam Injection Through Spargers Into a Large Water Pool2016Conference paper (Refereed)
    Abstract [en]

    The pressure suppression pool of a Boiling Water Reactor (BWR) is designed to protect the containment from over pressure by condensing steam. Under certain steam injection conditions, thermal stratification can develop in the pool and significantly reduce its pressure suppression capacity. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers, which are multi-hole injection pipes connecting the main steam lines to the wetwell pool. The aim of the model is to predict the global pool behavior. Effective Heat and Momentum Sources (EHS/EMS) approach is used to model time averaged effects of small scale direct contact condensation phenomena on the large scale pool circulation. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments was done to address the most important physical phenomena that can occur in plant scale. The results show that the global pool behavior can be predicted using the Standard Gradient Diffusion Hypothesis (SGDH) in k-Omega turbulence model.

  • 23.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Modeling of Thermal Stratification and Mixing in a Pressure Suppression Pool Using GOTHIC2016Conference paper (Refereed)
    Abstract [en]

    The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it can lead to higher containment pressures than in completely mixed conditions. The thermal hydraulic code of GOTHIC offers a very suitable platform to simulate the pool and containment behavior during a long term accident. However, for a computationally efficient code such as GOTHIC, direct contact condensation cannot be resolved accurately enough to obtain a good estimation of the momentum induced by the condensing steam, and thus, to predict the pool behaviour. In this paper, we present how to implement the previously validated Effective Heat Source (EHS) and Effective Momentum Source (EMS) models, developed for pool analysis during a steam injection, in GOTHIC. The implementation was done using control variables and Dynamically Linked Libraries (DLL). A time averaging model to minimize the effect of the numerical oscillations appearing in GOTHIC when steam is injected into the pool is also proposed.

  • 24.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Possibility of Air Ingress into a BWR Containment during a LOCA in case of Activation of Containment Venting System2014Conference paper (Refereed)
    Abstract [en]

    The pressure relief systems installed in BWRs protect the containment from overpressure in case of a Loss of Coolant Accident (LOCA). This paper analyzes the possibility of air ingress, which can cause hydrogen burn, through the rupture disks of the filtered and non-filtered venting systems. Two scenarios were considered: a LOCA without SBO (Station Blackout) and a LOCA with SBO. The thermal-hydraulic code GOTHIC® was used with 3D models of the drywell and wetwell of a Nordic-type BWR. In the LOCA event, we found no activation of the rupture disks within the considered transient simulation. Moreover, the containment spray ensured a low pressure in the drywell and induced a continuous mixing of the wetwell pool. In the LOCA with SBO event, the development of thermal stratification in the wetwell pool accelerated the pressure increase in the drywell, which led to activation of the rupture disk of the filtered venting system. However, no air ingress through the vent was found during the depressurization of the containment, and hence no risk of hydrogen burn under the given assumptions.

  • 25.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Scaling and CFD Modelling of the Pool Experiments with Spargers Performed in the PANDA Facility2016Conference paper (Refereed)
    Abstract [en]

    The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it reduces its cooling capability and leads to higher containment pressures than in completely mixed conditions. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers. The model provides the time averaged heat and momentum transferred from the steam condensation to the large scale pool circulation. Small scale phenomena such as direct contact condensation is not resolved, only its effect on the pool behaviour. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments, done to preserve the most important physical phenomena occurring in plant scale is also presented in the paper. The results show that the model is able to predict well the global pool behavior. However, flow instabilities were observed to induce a sudden mixing of the upper part of the stratified layer during the transition from the stratification to the mixing phases. This led to a faster erosion of the layer than in the experiment. Simulations done with 2D and 3D meshes and scale adaptive turbulence models were performed to clarify this issue and are presented in the paper.

  • 26.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Water Distribution in a Nordic BWR Containment During a LOCA2016In: 2016 International Congress on Advances in Nuclear Power Plants, ICAPP 2016, 2016Conference paper (Refereed)
    Abstract [en]

    During a main steam line break in a Boiling Water Reactor (BWR) the pressure suppression pool is used as a water source for the Emergency Core Cooling System (ECCS) and the Containment Spray (CS). These systems drain water from the pool through strainers, which are long perforated plates or cylinders submerged to a certain depth. Proper functioning of the ECCS and the CS must be ensured to maintain the water inventory in the vessel and to limit the containment pressure. However, if the liquid level in the suppression pool goes below the level of the strainers intake, the operators would be forced to stop their pumps. The liquid level in the suppression pool can be reduced when a significant fraction of ECCS and CS flow is relocated to the lower drywell. In this work, we use the thermal-hydraulic code GOTHIC to simulate the containment evolution during a main steam line break inside the biological shield. The containment volumes and their connections were modeled with 2D and 3D volumes. With this model, scenarios considering different operational conditions were assessed: (i) full capacity of all the safety systems, (ii) half capacity of all the safety systems, (iii) ECCS stops injecting water after a certain liquid level is restored in the vessel, and (iv) the pipes used to drain water from the suppression pool and flood the lower drywell are partially or totally clogged in different directions. The results showed that there is a risk of an early shut down of the ECCS and CS systems in the case of main steam line break inside the biological shield. It was observed that when the ECCS provided a continuous water injection into the vessel, the water spilled through the break into the biological shield flowed downwards driven by gravity and went directly into the lower drywell. This caused a fast decrease in the liquid level of the suppression pool, which led to an uncovery of the ECCS and CS strainers about 2000 s after the break. The activation at 1800 s of the flooding of the lower drywell led to a backward flow, from the lower drywell to the suppression pool, since at that time the liquid level in the suppression pool was lower than in the lower drywell. However, this backward flow was not enough to maintain the liquid level in the suppression pool, which continued to decrease. In the case where the pipes used for the flooding were clogged in the direction of the suppression pool, uncovery of the strainers was observed even earlier.

  • 27.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Risk analysis framework for severe accident mitigation strategy in nordic BWR: An approach to communication and decision making2017In: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, American Nuclear Society , 2017, p. 587-594Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWRs) employ ex-vessel debris cooling in a deep water pool. The success of the strategy requires (i) formation of a coolable porous debris bed; (ii) no energetic steam explosion that can threaten containment integrity. Both scenario (aleatory) and modeling (epistemic) uncertainties are important in the assessment of the failure risks. A consistent approach is necessary for the decision making on whether the strategy is sufficiently effective, or a modification of the SAM is necessary. Risk Oriented Accident Analysis Methodology (ROAAM+) is a tool for assessment of failure probability to enable robust decision making, insensitive to remaining uncertainty. Conditional containment failure probability is considered in this work as an indicator of severe accident management effectiveness for Nordic BWR. The ultimate goal of ROAAM+ application for Nordic BWR is to provide a scrutable background in order to achieve convergence of experts' opinions in decision making. The question is: if containment failure can be demonstrated as physically unreasonable, given severe accident management strategy and state-of-the-art knowledge? If inherent safety margins are large, then the answer to the question is positive and can be demonstrated through risk assessment with consistent conservative treatment of uncertainties and by improving, when necessary, knowledge and data. Otherwise, the risk management should be applied in order to increase margins and achieve the safety goal through modifications of the SAM (e.g. safety design, SAMGs, etc.). The challenge for a decision maker is to distinguish when collecting more knowledge and reduction of uncertainty in risk assessment or application of risk management with SAM modifications would be the most effective and efficient approach. In this work we demonstrate a conceptual approach for communication of ROAAM+ framework analysis results and provide an example of a decision support model. The results of the risk analysis are used in order to provide necessary insights on the conditions when suggested changes in the safety design are justified.

  • 28.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 125-141Article in journal (Refereed)
    Abstract [en]

    Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario parameters. Pattern analysis is employed in order to characterize typical behavior of core relocation transients. Clustering analysis is employed for grouping of different accident scenarios, which result in similar core relocation behavior and properties of the debris.

  • 29.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Scenario Grouping and Classification Methodology for Postprocessing of Data Generated by Integrated Deterministic-Probabilistic Safety Analysis2015In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 278638Article in journal (Refereed)
    Abstract [en]

    Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) combines deterministic model of a nuclear power plant with a method for exploration of the uncertainty space. Huge amount of data is generated in the process of such exploration. It is very difficult to "manually" process and extract from such data information that can be used by a decision maker for risk-informed characterization, understanding, and eventually decision making on improvement of the system safety and performance. Such understanding requires an approach for interpretation, grouping of similar scenario evolutions, and classification of the principal characteristics of the events that contribute to the risk. In this work, we develop an approach for classification and characterization of failure domains. The method is based on scenario grouping, clustering, and application of decision trees for characterization of the influence of timing and order of events. We demonstrate how the proposed approach is used to classify scenarios that are amenable to treatment with Boolean logic in classical Probabilistic Safety Assessment (PSA) from those where timing and order of events determine process evolution and eventually violation of safety criteria. The efficiency of the approach has been verified with application to the SARNET benchmark exercise on the effectiveness of hydrogen management in the containment.

  • 30.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Sensitivity analysis of debris properties in lower plenum of a Nordic BWR2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 374-382Article in journal (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. The melt released from the vessel is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. Melt release conditions are recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena, subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures (such as instrumentation guide tubes IGTs, control rod guide tubes CRGTs), vessel failure and melt release. This work is focused on the evaluation of uncertainty in core degradation progression and its effect on the resultant properties of relocated debris in lower plenum of Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize the range of possible debris properties in lower plenum and its sensitivity towards different modelling parameters, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the risk assessment framework.

  • 31.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Krčál, Pavel
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Ranlöf, Lisa
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Bäckström, Ola
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Adolfsson, Yvonne
    Lloyd’s Register Consulting, Stockholm, Sweden.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An approach to joint application of integrated deterministic-probabilistic safety analysis and PSA Level 2 to severe accident issues in Nordic BWRs2016In: PSAM 2016 - 13th International Conference on Probabilistic Safety Assessment and Management2017, International Association for Probabilistic Safety Assessment and Management (IAPSAM), 2016Conference paper (Refereed)
    Abstract [en]

    In this paper we outline a conceptual approach for combined use of Probabilistic Safety Assessment (PSA) and Integrated Deterministic-Probabilistic Safety Assessment (IDPSA), considering Nordic Boiling Water Reactor (BWR) severe accident issues (specifically ex-vessel steam explosion and debris bed coolability) for illustration.

    We describe a conceptual approach based on post processing of the results generated by IDPSA to update the “static” Boolean structures in the standard PSA representation. The challenge in the evaluation is to retain the failure combinations from the PSA to allow for component importance evaluation, to be able to perform the calculations in a reasonable time frame and to use all relevant information from the IDPSA results.

    We discuss the approaches for determination of the event space (for IDPSA analysis) which is consistent with PSA damage states from PSA-L1 and L2. We also discuss application of post processing approach for analysis of huge amount of data generated in the process of uncertainty space exploration, which is difficult to use directly in decision making process including incorporation of such data into PSA framework, to update structures of “static” Boolean structures in standard PSA. Using data post processing approach we can significantly reduce amount of information which represents results from IDPSA analysis.

  • 32.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Approach to Grouping and Classification of Scenarios in Integrated Deterministic-Probabilistic Safety Analysis2014In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper (Other academic)
    Abstract [en]

    Integrated Deterministic Probabilistic Safety Assessment (IDPSA) methodologies aim to achieve completeness and consistency of the analysis. However, for the purpose of risk informed decision making it is often insufficient to merely calculate a quantitative value for the risk and its associated uncertainties. IDPSA combines deterministic model of a nuclear power plant with a method for exploration of the uncertainty space. Huge amount of data is generated usually in the process of such exploration. It is very difficult to "manually" process and extract from such data information that can be used by a decision maker for risk-informed characterization and eventually improvement of the system safety and performance. Such understanding requires an approach to the interpretation, grouping of similar scenario evolutions, and classification of the principal characteristics of the events that contribute to the risk. In this work we develop an approach for classification and characterization of failure domains (domains of uncertain parameters where critical system parameters exceed safety thresholds). The method is based on scenario grouping and clustering with application of decision trees for characterization of the influence of timing and order of the events

  • 33.
    Galusin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of Core Relocation Surrogate Model for Prediction of Debris Properties in Lower Plenum of a Nordic BWR2016In: NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Gyeongju, Korea, October 9-13, 2016. Paper N11P1234, NUTHOS-11 , 2016, article id N11P1234Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. A consistent treatment of these uncertainties requires Integrated Deterministic Probabilistic Safety Analysis (IDPSA). We employ the concepts and approaches described in Risk Oriented Accident Analysis Methodology (ROAAM) for development of a probabilistic framework (ROAAM+) that is based on extensive uncertainty and sensitivity analysis in risk quantification. Direct application of such fine-resolution models for extensive sensitivity and uncertainty analysis is often unaffordable. We use “surrogate models” (SMs) that provide computationally efficient approximations for the FMs. In this work we demonstrate an approach to the development of Core relocation SM based on the MELCOR code as the full model (FM). We discuss the development of the database of the FM solutions, data mining and post-processing of the results for SM development. Extensive sensitivity and uncertainty analysis is carried out using the FM and implications of the analysis are discussed in detail. We demonstrate how the connection between different stages of severe accident progression is made in ROAAM+ framework for Nordic BWRs.

  • 34.
    Goronovski, Andrei
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi-Thanh
    Effect of Corium Non-homogeneity on Nordic BWR Vessel Failure Mode and Timing2015Conference paper (Refereed)
    Abstract [en]

    Corium melt fragmentation and cooling in a deep pool of water under reactor pressure vessel are employed as severe accident mitigation strategy in a Nordic-type BWR. Core debris relocated to the lower head inflict significant thermal and mechanical loads on the vessel structures. The mode and timing of the vessel failure, mass and superheat of the ejected melt determine ex-vessel accident progression and risks of steam explosion and formation of a non-coolable debris bed. In this work we consider the effect of in-vessel debris non-homogeneity on the mode of vessel failure. The heat-up, re-melting, melt pool formation, and heat transfer of the debris bed are predicted with the Phase-change Effective Convectivity Model (PECM) implemented in FLUENT® code. Then the obtained thermal load on the vessel wall and structures is used as boundary conditions for a thermo-structural analysis of the BWR lower head using the ANSYS® code. In this paper, a corium debris bed is considered inside vessel lower head inducing thermal load on the wall and structures. The debris bed thermal properties axial distribution is taken as a function of material composition, which is extracted from MELCOR® simulations of core failure and debris bed formation inside the lower plenum. A flat and a concave configuration of the debris bed are considered and results of simulations are compared with those for a homogenous debris bed of the same mass-averaged thermal properties.

  • 35.
    Goronovski, Andrei
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    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.
    The Effect of Internal Pressure and Debris Bed Thermal Properties on BWR Vessel Lower Head Failure and Timing2013In: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Conference paper (Refereed)
  • 36.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 9, p. 7222-7235Conference paper (Refereed)
    Abstract [en]

    Severe accident mitigation strategies in Nordic boiling water reactors (BWRs) employ core melt cooling in a deep pool of water under the reactor pressure vessel. Corium melt released from the vessel is expected to fragment, solidify and form a porous debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. Significant aleatory and epistemic uncertainties exist in accident scenarios, melt release conditions, and modeling of steam explosion phenomena. Assessment of the risk of ex-vessel steam explosion requires application of the Integrated Deterministic Probabilistic Safety Analysis (IDPSA). IDPSA is a computationally demanding task which makes unfeasible direct application of Fuel-Coolant Interaction codes. The goal of the current work is to develop a Surrogate Model (SM) of the Texas-V code and demonstrate its application to the analysis of explosion impact in the Nordic BWR. The SM should be computationally affordable for IDPSA analysis. We focus on prediction of the steam explosion loads in a reference Nordic BWR design assuming a scenario of coherent corium jet release into a deep water pool. We start with the review of the Texas-V sub-models in order to identify a list of parameters to be considered in implementation of the SM. We demonstrate that Texas-V exhibits chaotic response in terms of the explosion impulse as a function of the triggering time and introduce a statistical representation of the explosion impulse for given melt release conditions and arbitrary triggering time. We demonstrate that characteristics of the distribution are well-posed. We then separate out the essential portion of modelling uncertainty by identification of the most influential uncertain parameters using sensitivity analysis. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in FCI modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a Nordic BWR. A database of Texas-V solutions is generated and used for the development of the SM. Performance, predictive capability and application of the SM to risk analysis are discussed in detail.

  • 37.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    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 surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 311-327Article in journal (Refereed)
    Abstract [en]

    Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks' (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

  • 38.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity Study of Steam Explosion Characteristics to Uncertain Input Parameters Using TEXAS-V Code2014In: NUTHOS10, Paper-1293, Okinawa, Japan, 2014, Atomic Energy Society of Japan , 2014Conference paper (Refereed)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. Corium melt is expected to fragment, solidify and form a debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. There are many factors and parameters that could be considered for prediction of the fuel-coolant interaction (FCI) energetics, but it is not clear which of them are the most influential and should be addressed in risk analysis. The goal of this work is to assess importance of different uncertain input parameters used in FCI code TEXAS-V for prediction of the steam explosion energetics. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a reference design of a Nordic BWR. Sensitivity analysis with Morris method is implemented using coupled TEXAS-V and DAKOTA codes. In total 12 input parameters were studied and 2 melt release scenarios were considered. Each scenario is based on 60,000 of TEXAS-V runs. Sensitivity study identified the most influential input parameters, and those which have no statistically significant effect on the explosion energetics. Details of approach to robust usage of TEXAS-V input, statistical enveloping of TEXAS-V output and interpretation of the results are discussed in the paper. We also provide probability density function (PDF) of steam explosion impulse estimated using TEXAS-V for reference Nordic BWR. It can be used for assessment of the uncertainty ranges of steam explosion loads for given ranges of input parameters.

  • 39.
    Grishchenko, Dmitry
    et al.
    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.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes2015In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, p. 144-153Article in journal (Refereed)
    Abstract [en]

    Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  • 40.
    Grishchenko, Dmitry
    et al.
    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.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Design and Commissioning Tests of the TALL-3D Experimental Facility for Validation of Coupled STH and CFD Codes2014Conference paper (Refereed)
    Abstract [en]

    Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  • 41.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinova, Valtyna
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Insight into steam explosion in stratified melt-coolant configuration2013In: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 2013Conference paper (Refereed)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident  mitigation  strategy.  When  vessel  breach  is  large  and  water  pool  is shallow,  released  corium  melt  can  reach  containment  floor  in  liquid  form  and spread under water creating a stratified configuration of melt covered by coolant. Steam  explosion  in  such  stratified  configuration  was  long  believed  as  of secondary importance for reactor safety because it was assumed that considerable mass of melt cannot be premixed with the coolant. In this work we revisit these assumptions  using  recent  experimental  observations  from  the  stratified  steam explosion tests  in  PULiMS  facility.  We  demonstrate  that  (i)  considerable  melt-coolant premixing layer can be formed in the stratified configuration with high temperature  melts,  (ii)  mechanism  responsible  for  the  premixing  is  apparently more  efficient  than  previously  assumed  Rayleigh-Taylor  or  Kelvin-Helmholtz instabilities.  We  also  provide  data  on  measured  and  estimated  impulses, energetics  of  steam  explosion,  and  resulting  thermal  to  mechanical  energy conversion ratios. 

  • 42.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Mickus, Ignas
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    Experimental activities report on TALL-3D2015Report (Other academic)
  • 43.
    Hua, Li
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of Effective Momentum and Heat Flux Models for Stratification and Mixing in a Water Pool2013Report (Other academic)
    Abstract [en]

    The pressure suppression pool is the most important feature of the pressure suppression system in a Boiling Water Reactor (BWR) that acts primarily as a passive heat sink during a loss of coolant accident (LOCA) or when the reactor is isolated from the main heat sink. The steam injection into the pool through the blowdown pipes can lead to short term dynamic phenomena and long term thermal transient in the pool. The development of thermal stratification or mixing in the pool is a transient phenomenon that can influence the pool's pressure suppression capacity. Different condensation regimes depending on the pool's bulk temperature and steam flow rates determine the onset of thermal stratification or erosion of stratified layers. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (i) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (ii) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale modeling, and (c) validate our proposed models. The data from PPOOLEX STR-06, STR-09 and STR-10 tests are used for validation of the EHS and EMS models in this work. We found that estimations of the amplitude and frequency based on available experimental data from PPOOLEX experiments STR-06, STR-09, and STR-10 have too large uncertainties due to poor space and time resolution of the temperature measurements in the blowdown pipe. Nevertheless, the results demonstrated that simulations with variable effective momentum which is selected within the experimental uncertainty have provided reasonable agreement with test data on transient temperature distribution in the pool. In order to reduce uncertainty in both experimental data and EHS/EMS modeling, additional tests and modifications to the experimental procedures and measurements system in the PPOOLEX facility were proposed. Pre-test simulations were performed to aid in determining experimental conditions and procedures. Then, a new series of PPOOLEX experimental tests were carried out. A validation of EHS/EMS models against MIX-01 test is presented in this report. The results show that the clearing phase predicted with 3D drywell can match the experiment very well. The thermal stratification and mixing in MIX-01 is also well predicted in the simulation.

  • 44.
    Hultgren, Ante
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gallego-Marcos, Ignacio
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Simulation of Large Scale Erosion of a Stratified Helium Layer by a Vertical Air Jet using the GOTHIC Code2014Conference paper (Refereed)
    Abstract [en]

    In case of a severe core degradation in a Light Water Reactor (LWR), significant amount of hydrogen can be produced posing a risk of hydrogen burning and detonation. Reliable prediction of hydrogen build-up, stratification, and mixing in the containment is of paramount importance since the phenomena affect hydrogen distribution in the containment. In this paper, we present a modeling approach using the GOTHIC code. The simulation results were compared against experimental data from the ST1-7 experiment performed in the PANDA facility at the Paul Scherrer Institute (PSI). The ST1-7 experiment consists of an air jet impingement onto a stratified helium layer. The modelling approach uses coupled volumes to introduce in each region of the computational domain (i) adequate mesh resolutions to resolve the gradients of the flow and (ii) appropriate turbulence models in order to resolve locally dominant flow structures. With the adaptive mesh, only about 7400 cells for the 2 PANDA vessels (4 m diameter by 8 m in height cylinders with an interconnecting pipe) is enough to provide reasonably accurate results. We found that using the k-epsilon standard model for the jet region and the mixing length model for the rest of the domain, has provided remarkably good agreement with the experimental data. The erosion of the helium stratified layer before and after the air injection is discussed in detail.

  • 45.
    Isaev, S.A.
    et al.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Kudinov, Pavel
    Dnepropetrovsk State University.
    Kudravtsev, N.A.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Pyshnyi, I.A.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Numerical analysis of the jet-vortex pattern of flow in a rectangular trench2003In: Journal of Engineering Physics and Thermophysics, ISSN 1062-0125, E-ISSN 1573-871X, Vol. 76, no 2, p. 257-265Article in journal (Refereed)
    Abstract [en]

    Numerical simulation of the vortex structure of three-dimensional laminar flow in a rectangular trench of square cross section has been carried out on the basis of the finite-volume solution of steady-state Navier-Stokes equations.

  • 46. Jeltsov, Marti
    et al.
    Cadinu, F.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, P.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Approach to Validation of Coupled CFD and System Thermal-Hydraulics Codes.2011Conference paper (Refereed)
  • 47.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed)
  • 48. Jeltsov, Marti
    et al.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Simulation of a Steam Bubble Transport in the Primary System of the Pool Type Lead Cooled Fast Reactors2011In:  , 2011Conference paper (Refereed)
  • 49.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of multi-scale simulation methodology for analysis of heavy liquid metal thermal hydraulics with coupled STH and CFD codes2012In: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Conference paper (Refereed)
  • 50.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of TALL-3D Facility Design for Validation of Coupled STH and CFD Codes2012In: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Conference paper (Refereed)
1234 1 - 50 of 173
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