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  • 1.
    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.

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

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

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

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

  • 4.
    Do-Quang, Minh
    et al.
    KTH, School of Engineering Sciences (SCI), Centres, Linné Flow Center, FLOW.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Centres, Linné Flow Center, FLOW.
    Singer-Loginova, Irina
    KTH, School of Engineering Sciences (SCI), Centres, Linné Flow Center, FLOW.
    Amberg, Gustav
    KTH, School of Engineering Sciences (SCI), Centres, Linné Flow Center, FLOW.
    Parallel adaptive computation of some time-dependent materials-related microstructural problems2007In: Bulletin of the Polish Academy of Sciences: Technical Sciences, ISSN 0239-7528, Vol. 55, no 2, p. 229-237Article in journal (Refereed)
    Abstract [en]

    Some materials-related microstructural problems calculated using the phase-field method are presented. It is well known that the phase field method requires mesh resolution of a diffuse interface. This makes the use of mesh adaptivity essential especially for fast evolving interfaces and other transient problems. Complex problems in 3D are also computationally challenging so that parallel computations are considered necessary. In this paper, a parallel adaptive finite element scheme is proposed. The scheme keeps the level of node and edge for 2D and level of node and face for 3D instead of the complete history of refinements to facilitate derefinement. The information is local and exchange of information is minimized and also less memory is used. The parallel adaptive algorithms that run on distributed memory machines are implemented in the numerical simulation of dendritic growth and capillary-driven flows.

  • 5.
    Fichot, F.
    et al.
    IRSN, Cadarache, France. illanueva, W.; Bechta, S..
    Carenini, L.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A revised methodology to assess in-vessel retention strategy for high-power reactors2018In: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7, The American Society of Mechanical Engineers , 2018, Vol. 7Conference paper (Refereed)
    Abstract [en]

    The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the "3-layers" configuration, where the "focusing effect" may cause higher heat fluxes than in steady-state (due to transient "thin" metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 W/m(2)) whereas the first type provides the lowest heat fluxes (around 500 kW/m(2)) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes can reach, in many cases, values which are well above 1 MW/m(2). This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on "yield stress" (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed. The current approach followed by many experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena, e.g. associated with molten pool transient behaviour, and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion is more straightforward to be used in a deterministic approach.

  • 6.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kapulla, Ralf
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paranjape, Sidharth
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paladino, Domenico
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Laine, Jani
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Puustinen, Markku
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Rasanen, Antti
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Pyy, Lauri
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Kotro, Eetu
    Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 347, p. 67-85Article in journal (Refereed)
    Abstract [en]

    Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Generation III/III+ Pressurized Water Reactors (PWR) to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models are implemented in ANSYS Fluent 17.0 Computational Fluid Dynamics (CFD) code and calibrated against the spargers experiments performed in the PPOOLEX and PANDA facilities, analysed by the authors in Gallego-Marcos et al. (2018b). CFD modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the effective momentum showed an inverse proportionality to the sub-cooling. Differences between the effective momentum calibrated for PPOOLEX and PANDA are discussed. Analysis of the calculated flow above the cold stratified layer showed that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.

  • 7.
    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.

  • 8.
    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.

  • 9.
    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.

  • 10.
    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.

  • 11.
    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.

  • 12.
    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.

  • 13.
    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.

  • 14.
    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)
  • 15.
    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.

  • 16.
    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.

  • 17. Hotta, A.
    et al.
    Akiba, M.
    Morita, A.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Applied Physics, Nanostructure Physics. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Centres, Linné Flow Center, FLOW. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    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.
    Hoseyni, Seyed Mohsen
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sköld, Per
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Matsumoto, T.
    Sugiyama, T.
    Buck, M.
    Experimental and Analytical Investigation of Formation and Cooling Phenomena in High Temperature Debris Bed2019In: Journal of Nuclear Science and Technology, ISSN 0022-3131, E-ISSN 1881-1248Article in journal (Refereed)
    Abstract [en]

    Key phenomena in the cooling states of underwater debris beds were classified based on the premise that a target debris bed has a complicated geometry, nonhomogeneous porosity, and volumetric heat. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. Based on these classifications, the modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three tests, DEFOR-A, PULiMS, and REMCOD were carried in six phases to extend the existing database for validating implemented models. Up to Phase-5, the main part of these tests has been completed and the test plan has been modified from the original one due to occurrences of unforeseeable phenomena and changes in test procedures. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. Also, it tries to timely report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the shallow pool PULiMS tests.

  • 18.
    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.

  • 19.
    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.

  • 20. 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)
  • 21.
    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)
  • 22.
    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)
  • 23.
    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.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of a Domain Overlapping Coupling Methodology for STH/CFD Analysis of Heavy Liquid Metal Thermal-hydraulics2013In: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Conference paper (Refereed)
  • 24.
    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.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of domain overlapping STH/CFD coupling approach for analysis of heavy liquid metal thermal hydraulics in TALL-3D experiment2012Conference paper (Refereed)
  • 25.
    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.
    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.
    Validation of a CFD Code Star-CCM+ for Liquid Lead-Bismuth Eutectic Thermal-Hydraulics Using TALL-3D Experiment2014Conference paper (Refereed)
    Abstract [en]

    The engineering design, performance analysis and safety assessment of Generation IV heavy liquid metal cooled nuclear reactors calls for advanced and qualified numerical tools. These tools need to be qualified before used in decision making process. Computational Fluid Dynamics (CFD) codes provide detailed means for thermal-hydraulics analysis of pool-type nuclear reactors. This paper describes modeling of a forced to natural flow experiment in TALL-3D experimental facility using a commercial CFD code Star-CCM+. TALL-3D facility is 7 meters high LBE loop with two parallel hot legs and a cold leg. One of the hot legs accommodates the 3D test section, a cylindrical pool where the multi-dimensional flow conditions vary between thermal mixing and stratification depending on the mass flow rate and the power of the heater surrounding the pool. The pool outlet temperature which affects the natural convection flow rates in the system is governed by the flow structure in the pool. Therefore, in order to predict the dynamics of the TALL-3D facility it is crucial to resolve the flow inside the 3D test section. Specifically designed measurement instrumentation set-up provides steady state and transient data for calibration and validation of numerical models. The validity of the CFD model is assessed by comparing the computational results to experimental results.

  • 26.
    Jeltsov, Marti
    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.
    Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor2015In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 190, no 1, p. 1-10Article in journal (Refereed)
    Abstract [en]

    Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.

  • 27.
    Jeltsov, Marti
    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.
    Risk of sloshing in the primary system of a lead cooled fast reactor2014In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper (Refereed)
    Abstract [en]

    Pool-type designs of Lead-cooled Fast Reactor (LFR) aim for commercial viability by simplified engineering solutions and passive safety systems. However, such designs carry the risks related to heavy coolant sloshing in case of seismic event. Sloshing can cause (i) structural damage due to fluid-structure interaction (FSI) and (ii) core damage due to void induced reactivity insertion or due to local heat transfer deterioration. The main goal of this study is to identify the domain of seismic excitation characteristics at the reactor vessel level that can lead to exceedance of the safety limits for structural integrity and core damage. Reference pool-type LFR design used in this study is the European Lead-cooled SYstem (ELSY). Liquid lead sloshing is analyzed with Computational Fluid Dynamics (CFD) method. Outcome of the analysis is divided in two parts. First, different modes of sloshing depending on seismic excitation are identified. These modes are characterized by wave shapes, loads on structures and entrapped void. In the second part we capitalize on the framework of Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) to quantify the risk. Specifically, statistical parameters pertaining to mechanical loads and void transport are quantified and combined with the deterministically obtained data about consequences.

  • 28.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics.
    Seismic sloshing effects in lead-cooled fast reactors2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 99-110Article in journal (Refereed)
    Abstract [en]

    Pool-type primary system can improve the economy of lead-cooled fast reactors. However, partially filled pool of heavy liquid metal poses safety concerns related to seismic loads. Violent sloshing during earthquake-initiated fluid-structure interaction can lead to structural failures, gas entrapment and potential core voiding. Seismic isolation systems can be used to reduce the structural stresses, but its effect on sloshing is not straightforward. This paper presents a numerical study of seismic sloshing in ELSY reactor. The purpose is to evaluate the effects of seismic isolation system on sloshing at different levels of earthquake. Sloshing is modeled using computational fluid dynamics with a volume of fluid free surface capturing model. Earthquake is simulated using synthetic seismic data produced in SILER project as a boundary condition. Simultaneous verification and validation of the numerical model using a dam break experiment is presented. The adverse resonance effect of seismic isolation system is demonstrated in terms of sloshing-induced hydrodynamic loads and gas entrapment. Effectiveness of seismic isolation system is discussed separately for design and beyond design seismic levels. Partitioning baffles are proposed as a potential mitigation measure in the design and their effect is analyzed.

  • 29.
    Jeltsov, Marti
    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.
    Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 328, p. 255-265Article in journal (Refereed)
    Abstract [en]

    Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.

  • 30.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Yakush, Sergey
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. AlbaNova University Center.
    Phung, Viet-Anh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Nam
    A framework for assessment of severe accident management effectiveness in Nordic BWR plants2014In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper (Refereed)
    Abstract [en]

    In the case of severe accident in Nordic boiling water reactors (BWR), core melt is poured into a deep pool of water located under the reactor. The severe accident management (SAM) strategy involves complex and coupled physical phenomena of melt-coolant-structure interactions sensitive to the transient accident scenarios. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) if corium debris bed is coolable, and (ii) potential for energetic steam explosion. The goal of this work is to develop a risk-oriented accident analysis framework for quantifying conditional threats to containment integrity for a Nordic-type BWR. The focus is on the process of refining the treatment and components of the framework to achieve (i) completeness, (ii) consistency, and (iii) transparency in the review of the analysis and its results. A two-level coarse-fine iterative refinement process is proposed. First, fine-resolution but computationally expensive methods are used in order to develop computationally efficient surrogate models. Second, coupled modular framework is developed connecting initial plant damage states with respective containment failure modes. Systematic statistical analysis is carried out to identify the needs for refinement of detailed methods, surrogate models, data and structure of the framework to reduce the uncertainty, and increase confidence and transparency in the risk assessment results.

  • 31.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Agglomeration and size distribution of debris in DEFOR-A experiments with Bi2O3-WO3 corium simulant melt2013In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 263, p. 284-295Article in journal (Refereed)
    Abstract [en]

    Flooding of lower drywell has been adopted as a cornerstone of severe accident management strategy in Nordic type Boiling Water Reactors (BWR). It is assumed that the melt ejected into a deep pool of water will fragment, quench and form a porous debris bed coolable by natural circulation. If debris bed is not coolable, then dryout and possibly re-melting of the debris can occur. Melt attack on the containment basemat can threaten containment integrity. Agglomeration of melt debris and formation of solid "cake" regions provide a negative impact on coolability of the porous debris bed. In this work we present results of experimental investigation on the fraction of agglomerated debris obtained in the process of hot binary oxidic melt pouring into a pool of water. The Debris Bed Formation and Agglomeration (DEFOR-A) experiments provide data about the effects of the pool depth and water subcooling, melt jet diameter, and initial melt superheat on the fraction of agglomerated debris. The data presents first systematic study of the debris agglomeration phenomena and facilitates understanding of underlying physics which is necessary for development and validation of computational codes to enable prediction of the debris bed coolability in different scenarios of melt release.

  • 32.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Tran, Chi Thanh
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    The DEFOR-A Experiment on Fraction of Agglomerated Debris as a Function of Water Pool Depth2010In: The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), 2010Conference paper (Refereed)
  • 33.
    Kudinov, Pavel
    et al.
    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.
    Fraction of Agglomerated Debris in Experiment on Pouring of High Temperature Melt in Water2010In: Proc. 8th International TopicalMeeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), 2010Conference paper (Refereed)
  • 34.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Condensation, Stratification and Mixing in aBWR Supression Pool2010Report (Other academic)
  • 35.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development and implementation of effective models in GOTHIC for the prediction of mixing and thermal stratification in a BWR pressure suppression pool2011In: Proceedings of the 2011 international congress on advances in nuclear power plants: ICAPP2011, American Nuclear Society, 2011Conference paper (Refereed)
    Abstract [en]

    As a passive safety system, the function of steam suppression pools in a BWR plant isparamount to the containment performance. The pressure suppression pool wasdesigned to have the capability as a heat sink to cool and condense steam releasedfrom the core vessel and/or main steam line during loss of coolant accident (LOCA)or opening of safety relief valve in normal operation of BWRs. For the case of smallflow rates of steam influx, thermal stratification could develop on the part aboveblowdown pipe exit and significantly impede the pool’s pressure suppression capacity.During a steam blowdown, the steam condenses rapidly in the pool and the hotcondensate rises in a narrow plume above steam injection plane and spreads into athin layer at the pool’s free surface. The increasing temperature of the surfacedefining the steam partial pressure in the vapor space causes increment ofcontainment pressure. Once steam flow rate increases significantly, momentumintroduced by the steam injection and/or periodic expansion and shrink of large steambubbles due to direct contact condensation can destroy stratified layers and lead tomixing of the pool water.Accurate and computationally efficient prediction of the pool thermal-hydraulics inthe scenarios with transition between thermal stratification and mixing, presents acomputational challenge. Lumped parameter codes have no capability to predicttemperature distribution of water pool during thermal stratification development.Scaling approaches and 1D codes generally have a lack of ability to capture historyeffects in complex plant transients, while high-order-accurate CFD (RANS, LES)methods are not practical due to excessive computing power needed to calculate 3Dhigh-Rayleigh-number natural circulation flow in long transients.In this paper we discuss a middle ground approach to modeling of mixing and thermalstratification development during steam injection in a tank of water. The approachemploys GOTHIC containment code as a computational vehicle which has features ofboth 1D/lumped parameter models for description of plant thermal hydraulic systemand CFD-like models for prediction of distributed parameters in the pool. The effectof steam injection on the mixing and stratification is provided by the effective heatsource (EHS) model and the effective momentum source (EMS) model. Proposedmodels are based on the experimental observations and analysis which suggest thatthe heat flux through the blowdown pipe and the momentum out of the pipe outlet aretwo driving factors which affect stratification and mixing process. The EMS modelprovides the effect of steam injection in terms of momentum responsible forestablishment of large scale circulation in the pool, while the aim of the EHS model isto simulate the thermal effect of steam injection by the effective heat transferred to thepool. The POOLEX experimental facility (Lappeenranta University of Technology inFinland) data is used to develop and validate the EHS and EMS models. Thecomparison of simulation results to experimental data are discussed in the paper.

  • 36.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development and Validation of Effective Models for Simulation of Stratification and Mixing Phenomena in a Pool of Water2011Report (Other academic)
    Abstract [en]

    This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. The pressure suppression pool was designed to have the capability as a heat sink to cool and condense steam released from the core vessel and/or main steam line during loss of coolant accident (LOCA) or opening of safety relief valve in normal operation of BWRs. For the case of small flow rates of steam influx, thermal stratification could develop on the part above the blowdown pipe exit and significantly impede the pool's pressure suppression capacity. Once steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and collapse of large steam bubbles due to direct contact condensation can destroy stratified layers and lead to mixing of the pool water. We use CFD-like model of the general purpose thermal-hydraulic code GOTHIC for addressing the issues of stratification and mixing in the pool. In the previous works we have demonstrated that accurate and computationally efficient prediction of the pool thermal-hydraulics in the scenarios with transition between thermal stratification and mixing, presents a computational challenge. The reason is that direct contact condensation phenomena, which drive oscillatory motion of the water in the blowdown pipes, are difficult to simulate with original GOTHIC models because of appearance of artificial oscillations due to numerical disturbances. To resolve this problem we propose to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) model and the Effective Momentum Source (EMS) model. We use POOLEX/PPOOLEX experiment (Lappeenranta University of Technology in Finland), in order 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) to validate proposed models. Results obtained with the EHS model shows that GOTHIC can predict development of thermal stratification in the pool if adequate grid resolution is provided. An equation for the effective momentum is proposed based on feasibility studies of the EMS model and analysis of the measured data in the test with chugging regime of steam injection. An experiment with higher resolution in space and time of oscillatory flow inside the blowdown pipe is highly desirable to uniquely determine model coefficients. Implementation of EHS/EMS model in GOTHIC and their validation against new PPOOLEX experiment is underway.

  • 37.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Modeling of Condensation,Stratification, and Mixing Phenomena in a Pool of Water2010Report (Other academic)
    Abstract [en]

    This work pertains to the research program on Containment Thermal-Hydraulicsat KTH. The objective is to evaluate and improve performance of methods, whichare used to analyze thermal-hydraulics of steam suppression pools in a BWRplant under different abnormal transient and accident conditions. As a passivesafety system, the function of steam pressure suppression pools is paramount tothe containment performance. In the present work, the focus is on apparentlybenign but intricate and potentially risk-significant scenarios in which thermalstratification could significantly impede the pool’s pressure suppression capacity.For the case of small flow rates of steam influx, the steam condenses rapidly inthe pool and the hot condensate rises in a narrow plume above the steam injection plane and spreads into a thin layer at the pool’s free surface. When thesteam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and shrink of large steam bubbles due to directcontact condensation can cause breakdown of the stratified layers and lead tomixing of the pool water. Accurate prediction of the pool thermal-hydraulics insuch scenarios presents a computational challenge. Lumped-parameter modelshave no capability to predict temperature distribution of water pool during thermalstratification development. While high-order-accurate CFD (RANS, LES) methodsare not practical due to excessive computing power needed to calculate 3D highRayleigh-number natural circulation flow in long transients. In the present work, amiddle-ground approach is used, namely CFD-like model of the general purposethermal-hydraulic code GOTHIC. Each cell of 3D GOTHIC grid uses lumped parameter volume type closures for modeling of various heat and mass transferprocesses at subgrid scale. We use GOTHIC to simulate POOLEX/PPOOLEXexperiment, in order to (a) quantify errors due to GOTHIC’s physical models andnumerical schemes, and (b) propose necessary improvements in GOTHIC subgrid scale modeling. The study performed on thermal stratification in a water poolindicates that GOTHIC CFD-like model is fit for reactor applications in complexfluid-physics scenarios that avoids both over-simplification (as in single lumpedparameter model) and over-complication (as in CFD models). However, simulation of direct steam injection into a subcooled pool cannot be predicted reliablywith the existing models. Thus we develop “effective heat source” and “effectivemomentum” approaches, and provide feasibility study for the prediction of thermalstratification and mixing in a BWR pressure suppression pool. The results areencouraging and further activity on the development and implementation of theproposed models in GOTHIC is currently underway.

  • 38.
    Li, Hua
    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.
    Approach and Development of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water2014In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, p. 108782-Article in journal (Refereed)
    Abstract [en]

    Steam venting and condensation in a large pool of water can lead to either thermal stratification or thermal mixing. In a pressure suppression pool (PSP) of a boiling water reactor (BWR), consistent thermal mixing maximizes the capacity of the pool while the development of thermal stratification can reduce the steam condensation capacity of the pool which in turn can lead to pressure increase in the containment and thereafter the consequences can be severe. Advanced modeling and simulation of direct contact condensation in large systems remain a challenge as evident in commercial and research codes mainly due to small time-steps necessary to resolve contact condensation in long transients. In this work, effective models, namely, the effective heat source (EHS) and effective momentum source (EMS) models, are proposed to model and simulate thermal stratification and mixing during a steam injection into a large pool of water. Specifically, the EHS/EMS models are developed for steam injection through a single vertical pipe submerged in a pool under two condensation regimes: complete condensation inside the pipe and chugging. These models are computationally efficient since small scale behaviors are not resolved but their integral effect on the large scale flow structure in the pool is taken into account.

  • 39.
    Li, Hua
    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.
    Effective Models for Prediction of Stratification and Mixing Phenomena in a BWR Supression Pool2012Report (Other academic)
  • 40.
    Li, Hua
    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.
    Effective Momentum and Heat Flux Models for Simulation of Stratification and Mixing in a Large Pool of Water2012Report (Other academic)
    Abstract [en]

    Performance of a boiling water reactor (BWR) containment is mostly determined by reliable operation of pressure suppression pool which serves as a heat sink to cool and condense steam released from the core vessel. Thermal stratification in the pool can significantly impede the pool's pressure suppression capacity. A source of momentum is required in order to break stratification and mix the pool. It is important to have reliable prediction of transient development of stratification and mixing in the pool in different regimes of steam injection. 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 (1) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (2) 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 POOLEX/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. Specifically the data from POOLEX STB-21 and PPOOLEX STR-03 and STR-04 tests are used for validation of the EHS and EMS models in this work. We show that the uncertainty in model prediction is comparable with the uncertainty in the experiments. The capability of the EHS/EMS model to predict thermal stratification and mixing in a plant scale pressure suppression pool is demonstrated. Finally, a new series of PPOOLEX experimental tests is proposed to reduce experimental uncertainty and to validate more accurately the sub-models used in the EMS model.

  • 41.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Puustinen, M.
    Laine, J.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal stratification and mixing in a suppression pool induced by direct steam injection2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 111, p. 487-498Article in journal (Refereed)
    Abstract [en]

    An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool. 

  • 42.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Puustinen, Markku
    Laine, Jani
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water2014In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, p. 752597-Article in journal (Refereed)
    Abstract [en]

    The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to predict the development of thermal stratification and mixing during a steam injection into a large pool of water. These effective models are implemented in GOTHIC software and validated against the POOLEX STB-20 and STB-21 tests and the PPOOLEX MIX-01 test. First, the EHS model is validated against STB-20 test which shows the development of thermal stratification. Different numerical schemes and grid resolutions have been tested. A 48x114 grid with second order scheme is sufficient to capture the vertical temperature distribution in the pool. Next, the EHS and EMS models are validated against STB-21 test. Effective momentum is estimated based on the water level oscillations in the blowdown pipe. An effective momentum selected within the experimental measurement uncertainty can reproduce the mixing details. Finally, the EHS-EMS models are validated against MIX-01 test which has improved space and time resolution of temperature measurements inside the blowdown pipe. Excellent agreement in averaged pool temperature and water level in the pool between the experiment and simulation has been achieved. The development of thermal stratification in the pool is also well captured in the simulation as well as the thermal behavior of the pool during the mixing phase.

  • 43.
    Ma, Weimin
    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.
    Miassodoev, A.
    Gaus-Liu, X.
    Hoffmann, M.
    Jaeckel, B.
    Birchley, J.
    Matejovic, P.
    Barnak, M.
    Progress in melt pool behavior and coolability in the lower head of a light water reactor2012Conference paper (Refereed)
  • 44.
    Mickus, Ignas
    et al.
    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.
    Vorobyev, Yuri
    Moscow Power Engineering Institute.
    Villanueva, Walter
    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 Physics Based Surrogate Model Development for Application with IDPSA2014Conference paper (Refereed)
    Abstract [en]

    Integrated Deterministic Probabilistic Safety Assessment (IDPSA) methodology is a powerful tool for identification of failure domains when both stochastic events and physical time dependent processes are important. Computational efficiency of deterministic models is one of the limiting factors for detailed exploration of the event space. Pool type designs of Generation IV heavy liquid metal cooled reactors introduce importance of capturing intricate 3D flow phenomena in safety analysis. Specifically mixing and stratification in 3D elements can affect efficiency of passive safety systems based on natural circulation. Conventional 1D System Thermal Hydraulics (STH) codes are incapable of predicting such complex 3D phenomena. Computational Fluid Dynamics (CFD) codes are too computationally expensive to be used for simulation of the whole reactor primary coolant system. One proposed solution is code coupling where all 1D components are simulated with STH and 3D components with CFD codes. However, modeling with coupled codes is still too time consuming to be used directly in IDPSA methodologies, which require thousands of simulations. The goal of this work is to develop a computationally efficient surrogate model (SM) which captures key physics of complex thermal hydraulic phenomena in the 3D elements and can be coupled with 1D STH codes instead of CFD. TALL-3D is a lead-bismuth eutectic thermal hydraulic loop which incorporates both 1D and 3D elements. Coupled STH-CFD simulations of TALL-3D typical transients (such as transition from forced to natural circulation) are used to calibrate the surrogate model parameters. Details of current implementation and limitations of the surrogate modeling are discussed in the paper in detail.

  • 45. Phung, Viet-Anh
    et al.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Raub, Sebastian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Goronovski, Andrei
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Koop, Kaspar
    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.
    Characteristics of debris in the lower head of a BWR in different severe accident scenarios2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, p. 359-370Article in journal (Refereed)
    Abstract [en]

    Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel-coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input can result in large changes in the final mass of debris. Typical ranges of the debris properties in different scenarios are discussed in detail.

  • 46.
    Phung, Viet-anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Raub, Sebastian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Goronovski, Andrei
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    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.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Prediction of Corium Debris Characteristics in Lower Plenum of a Nordic BWR in Different Accident Scenarios Using MELCOR Code2015In: 2015 International Congress on Advances in Nuclear Power Plants, Nice, France: ICAPP , 2015, , p. 11Conference paper (Refereed)
    Abstract [en]

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed configurations in the lower plenum.

  • 47.
    Torregrosa, Claudio
    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.
    Coupled 3D Thermo-mechanical Analysis of a Nordic BWR Vessel Failure and Timing2013In: Proceedings 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Conference paper (Refereed)
  • 48.
    Tran, Chi Thanh
    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.
    A Study on the Integral Effect of Corium Material Properties on Melt Pool Heat Transfer in a Boiling Water Reactor2011In: Proceedings 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), 2011Conference paper (Refereed)
  • 49.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Mechanics.
    Diffuse-Interface Simulations of Capillary Phenomena2007Doctoral thesis, comprehensive summary (Other scientific)
    Abstract [en]

    Fluid flows mainly driven by capillary forces are presented in this thesis. By means of modeling and simulations, interesting dynamics in capillary-driven flows are revealed such as coalescences, breakups, precursor films, flow instabilities, rapid spreading, rigid body motions, and reactive wetting.

    Diffuse-interface methods model a fluid interface as having a finite thickness endowed with physical properties such as surface tension. Two diffuse-interface models that are based on the free energy of the system are presented. The binary model, more specifically the coupled Navier-Stokes/Cahn-Hilliard equations, was used to study different two-phase flows including problems related to microfluidics. Numerical issues using this model have been addressed such as the need for mesh adaptivity and time-step restrictions. Moreover, the flexibility of this model to simulate 2D, axisymmetric, and 3D flows has been demonstrated.

    The factors affecting reproducibility of microdroplet depositions performed under a liquid medium are investigated. In the deposition procedure, sample solution is dispensed from the end of a capillary by the aid of a pressure pulse onto a substrate with pillar-shaped sample anchors. In both the experimental and numerical study it was shown that the deposited volume mainly depends on the capillary-substrate distance and anchor surface wettability. Furthermore, a critical equilibrium contact angle has been identified below which reproducible depositions are facilitated.

    The ternary model is developed for more complicated flows such as liquid phase sintering. With the introduction of a Gibbs energy functional, the governing equations are derived, consisting of convective concentration and phase-field equations which are coupled to the Navier-Stokes equations with surface tension forces. Arbitrary phase diagrams, surface energies, and typical dimensionless numbers are some input parameters into the model. Detailed analysis of the important capillary phenomena in liquid phase sintering such as reactive and nonreactive wetting and motion of two particles connected by a liquid bridge are presented. The dynamics of the wetting is found to match with a known hydrodynamic theory for spreading liquids. Factors affecting the equilibrium configuration of the particles such as equilibrium contact angles and volume ratios are also investigated.

  • 50.
    Villanueva, Walter
    et al.
    KTH, School of Engineering Sciences (SCI), Mechanics.
    Amberg, Gustav
    KTH, School of Engineering Sciences (SCI), Mechanics, Physicochemical Fluid Mechanics.
    Phase-field Simulations Of Free Boundary Microflow2006In: Proceedings of Second International Conference on Transport Phenomena in Micro and Nanodevices, 2006Conference paper (Refereed)
12 1 - 50 of 67
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