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  • 51.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi-Thanh
    Effect of Corium Non-homogeneity on Nordic BWR Vessel Failure Mode and Timing2015Konferensbidrag (Refereegranskat)
    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.

  • 52.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The Effect of Internal Pressure and Debris Bed Thermal Properties on BWR Vessel Lower Head Failure and Timing2013Ingår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferensbidrag (Refereegranskat)
  • 53.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR2015Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 9, s. 7222-7235Konferensbidrag (Refereegranskat)
    Abstract [en]

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

  • 54.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs2016Ingår i: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, s. 311-327Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 55.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sensitivity Study of Steam Explosion Characteristics to Uncertain Input Parameters Using TEXAS-V Code2014Ingår i: NUTHOS10, Paper-1293, Okinawa, Japan, 2014, Atomic Energy Society of Japan , 2014Konferensbidrag (Refereegranskat)
    Abstract [en]

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

  • 56.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR2017Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Konferensbidrag (Refereegranskat)
    Abstract [en]

    Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel in order to fragment and quench core melt and provide long term cooling of the debris. One of the risk factors associated with this accident management strategy is early failure of the containment due to steam explosion. Assessment of the risk is subject to significant epistemic and aleatory uncertainties in (i) modelling of steam explosion and (ii) scenarios of melt release from the vessel and water pool conditions. High computational efficiency of the models is required for such assessment. A surrogate model (SM) approach has been previously developed using artificial neural network and the database of Texas-V code solutions for steam explosion loads in the Nordic type BWRs. In this paper we extend our surrogate model to allow analysis of steam explosion in relatively shallow water pools (>2 m), address effects of melt emissivity and resolve more accurately variation of pressure in the drywell. We provide detailed comparison of metallic vs oxidic melt release scenarios, incorporate uncertainty of the SM into modelling and analyze the sensitivity of our results to SM uncertainty. We estimate risks of containment failure with non-reinforced and reinforced hatch door and demonstrate the effect of the surrogate model uncertainty on the results. We analyze the results and develop a simplified approach for decision making considering predicted failure probabilities, expected costs and scenario frequencies. 

  • 57.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes2015Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, s. 144-153Artikel i tidskrift (Refereegranskat)
    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.

  • 58.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Design and Commissioning Tests of the TALL-3D Experimental Facility for Validation of Coupled STH and CFD Codes2014Konferensbidrag (Refereegranskat)
    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.

  • 59.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinova, Valtyna
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Insight into steam explosion in stratified melt-coolant configuration2013Ingår i: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 2013Konferensbidrag (Refereegranskat)
    Abstract [en]

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

  • 60.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Mickus, Ignas
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    Experimental activities report on TALL-3D2015Rapport (Övrigt vetenskapligt)
  • 61.
    Hua, Li
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Validation of Effective Momentum and Heat Flux Models for Stratification and Mixing in a Water Pool2013Rapport (Övrigt vetenskapligt)
    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.

  • 62.
    Hultgren, Ante
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gallego-Marcos, Ignacio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Simulation of Large Scale Erosion of a Stratified Helium Layer by a Vertical Air Jet using the GOTHIC Code2014Konferensbidrag (Refereegranskat)
    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.

  • 63.
    Isaev, S.A.
    et al.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Kudinov, Pavel
    Dnepropetrovsk State University.
    Kudravtsev, N.A.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Pyshnyi, I.A.
    Acad. of Civil Aviation, Sankt-Peterburg, Russian Federation.
    Numerical analysis of the jet-vortex pattern of flow in a rectangular trench2003Ingår i: Journal of Engineering Physics and Thermophysics, ISSN 1062-0125, E-ISSN 1573-871X, Vol. 76, nr 2, s. 257-265Artikel i tidskrift (Refereegranskat)
    Abstract [en]

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

  • 64. Jeltsov, Marti
    et al.
    Cadinu, F.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, P.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An Approach to Validation of Coupled CFD and System Thermal-Hydraulics Codes.2011Konferensbidrag (Refereegranskat)
  • 65.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment2018Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikel i tidskrift (Refereegranskat)
  • 66.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment2019Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 341, s. 306-325Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).

  • 67. Jeltsov, Marti
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Simulation of a Steam Bubble Transport in the Primary System of the Pool Type Lead Cooled Fast Reactors2011Ingår i:  , 2011Konferensbidrag (Refereegranskat)
  • 68.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Development of multi-scale simulation methodology for analysis of heavy liquid metal thermal hydraulics with coupled STH and CFD codes2012Ingår i: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Konferensbidrag (Refereegranskat)
  • 69.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Development of TALL-3D Facility Design for Validation of Coupled STH and CFD Codes2012Ingår i: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Konferensbidrag (Refereegranskat)
  • 70.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Pre-test analysis of an LBE solidification experiment in TALL-3DIngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikel i tidskrift (Refereegranskat)
    Abstract [en]

    This paper presents the process of development of a solidification test section design for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop of prototypic height). Coolant solidification is a phenomenon of potential safety importance for liquid metal cooled fast reactors (LMFRs). Solidification, e.g. in case of excessive performance of the passive decay heat removal systems, can affect local heat transfer and even lead to partial or complete blockage of the coolant flow paths. This might lead to failure of decay heat removal function. In case of reduced flow circulation, the temperature of the coolant will increase, which might prevent complete blockage of the flow. Prediction of possible outcomes of such scenarios with complex interactions between local physical phenomena of solidification and system scale natural circulation behavior is subject to modelling (epistemic) uncertainty. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support experiment development. The aim of the experimental design is to satisfy requirements stemming from the process of qualification of the model that can be used for addressing the safety-related concerns. In this work we focus on two aspects: (i) design of solidification test section (STS) for investigation of local solidification phenomena of lead-bismuth eutectic (LBE), and (ii) effect of the STS on the system scale behavior of the experimental facility. We discuss selection of the STS characteristics and experimental test matrix using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.

  • 71.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Pre-test analysis of an LBE solidification experiment in TALL-3D2018Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 339, s. 21-38Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Coolant solidification is a phenomenon of potential safety importance for Liquid Metal Cooled Fast Reactors (LMFRs). Coolant solidification can affect local flow, heat transfer and lead to partial or complete blockage of the coolant flow paths jeopardizing decay heat removal function. It is also possible that reduced flow circulation may increase coolant temperature, counteract solidification and prevent complete blockage of the flow. Complex interactions between local physical phenomena of solidification and system scale natural circulation make modelling of solidification uncertain. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support of experiment development, specifically, design of a solidification test section and a test matrix for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop). The aim of the analysis is experimental design that satisfies requirements stemming from the process of model qualification. We focus on two aspects: (i) design of solidification test section (STS) for investigation of solidification phenomena in lead-bismuth eutectic (LBE), and (ii) effect of the STS pool on the system scale behavior of the TALL-3D facility. Selection of the STS characteristics and experimental test matrix is supported using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes. 

  • 72.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of a Domain Overlapping Coupling Methodology for STH/CFD Analysis of Heavy Liquid Metal Thermal-hydraulics2013Ingår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferensbidrag (Refereegranskat)
  • 73.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Development of domain overlapping STH/CFD coupling approach for analysis of heavy liquid metal thermal hydraulics in TALL-3D experiment2012Konferensbidrag (Refereegranskat)
  • 74.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of a CFD Code Star-CCM+ for Liquid Lead-Bismuth Eutectic Thermal-Hydraulics Using TALL-3D Experiment2014Konferensbidrag (Refereegranskat)
    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.

  • 75.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor2015Ingår i: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 190, nr 1, s. 1-10Artikel i tidskrift (Refereegranskat)
    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.

  • 76.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Risk of sloshing in the primary system of a lead cooled fast reactor2014Ingår i: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Konferensbidrag (Refereegranskat)
    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.

  • 77.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core2018Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 328, s. 255-265Artikel i tidskrift (Refereegranskat)
    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.

  • 78. Journeau, Christophe
    et al.
    Bouyer, Viviane
    Cassiaut-Louis, Nathalie
    Fouquart, Pascal
    Piluso, Pascal
    Ducros, Gerard
    Gosse, Stephane
    Gueneau, Christine
    Quaini, Andrea
    Fluhrer, Beatrix
    Miassoedov, Alexei
    Stuckert, Juri
    Steinbrueck, Martin
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Hozer, Zoltan
    Guba, Attila
    Manara, Dario
    Bottomley, David
    Fischer, Manfred
    Langrock, Gert
    Schmidt, Holger
    Kiselova, Monika
    Zdarek, Jiri
    SAFEST ROADMAP FOR CORIUM EXPERIMENTAL RESEARCH IN EUROPE2016Ingår i: PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 4, ASME Press, 2016, Vol. 4, artikel-id V004T14A019Konferensbidrag (Refereegranskat)
    Abstract [en]

    SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities deteimined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100500 kg) prototypic corium facility.

  • 79.
    Karbojian, Aram
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, Mikhail
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A scoping study of debris formation in DEFOR experimental facility2007Ingår i: Proceedings of the 15th International Conference on Nuclear Engineering (ICONE15), 2007Konferensbidrag (Refereegranskat)
  • 80.
    Karbojian, Aram
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A scoping study of debris bed formation in the DEFOR test facility2009Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, nr 9, s. 1653-1659Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Motivated to understand the processes which govern the formation and characteristics of a debris bed and hence its coolability during a postulated severe accident of a light water reactor, a new research program called DEFOR (DEbris FORmation) was initiated at the Royal Institute of Technology (KTH). This paper presents results obtained in scoping experiments conducted during an initial phase of the DEFOR program. The DEFOR-E test campaign is concerned with the DEFOR test facility commissioning and exploratory study of phenomena occurred during a debris bed formation. Binary oxide mixtures at different superheat temperatures were used as the corium melt simulants. The scoping experiments revealed the effect of water pool depth and subcooling, melt mass and material properties on the debris bed characteristics. Insights gained from the DEFOR-E test campaign help guide the scaling, design and operation of the subsequent experiments in the DEFOR program.

  • 81.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Experimental and Analytical Study of the Particulate Debris Bed Self-leveling2012Ingår i: The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), Kaohsiung, Taiwan, September 9-13, 2012, 2012Konferensbidrag (Refereegranskat)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel is employed as a severe accident (SA) mitigation strategy in several designs of light water reactors (LWR). Geometrical configuration of the debris bed is one of the factors which define if the decay heat can be removed from the debris bed by natural circulation. Boiling and two-phase flow inside the bed also serves as a source of mechanical energy which can reduce the height of the debris bed by so called “self-leveling” phenomenon. However, to be effective in providing a coolable geometrical configuration, self-leveling time scale has to be smaller than the time scale for drying out and onset of re-melting of the bed. This paper presents results of experimental and analytical studies concerning the self-leveling phenomenon. The goal of this work is to assess characteristic time scale of particulate debris spreading. In the experiments on the particulate debris spreading air injection at the bottom of the bed is used to simulate steam flow through the porous debris bed. A series of test have been carried out to study the influence of particles size and density, roughness of the spreading plate, gas flow rate etc. on particulate spreading. A semi-empirical model for predicting the spreading of particulate debris has been developed using experimental closures for debris mass flow rate as a function of local (i) angle of the bed and (ii) gas flux. The comparison between the model prediction and the experimental observations shows a good agreement.

  • 82.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Experiments and Characterization of the Two-Phase Flow Driven Particulate Debris Spreading in the Pool2014Ingår i: NUTHOS-10 / [ed] http://www.nuthos10.org/, Okinawa, Japan, 2014, s. 1257-Konferensbidrag (Refereegranskat)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase turbulent flows in the pool serve as a source of mechanical energy which can affect the initial geometry as well as dynamically change the shape of already formed debris bed. The main goal of this work is to provide experimental data on spreading of solid particles in the pool by large scale two-phase flow structures induced by gas injection from the bottom. These data are necessary for development and validation of predictive capabilities of computer codes allowing numerical modeling of the debris bed formation at prototypic severe accident conditions.  Results of a new series of PDS-P (Particulate Debris Spreading in the Pool) tests reported in this paper are for two types of tests: (i) the pure two-phase flows without particles and (ii) tests with particles. In both tests series, vapor flows in saturated water are simulated by air injection at the bottom of the facility. Experimental conditions such as gas-phase flow rate and particle properties (density, size etc.) are scaled to maintain relevancy to the prototypic accident conditions. The water pool is constructed as a rectangular tank. It has close to 2D geometry with fixed width (72 mm), variable length (up to 1.6 m) and allows water filling depth of up to 1 m. The variable pool length and depth allows formation of the different in size and pattern two-phase circulating flows. The average void fraction in the pool is determined by video recording and image processing. Particles are supplied from the top of the facility above the water surface. In the separate-effect studies of the influence of two-phase currents on particle trajectories and bed formation, low particle flow rate is required in order to minimize or completely exclude particle-particle interaction.

  • 83.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, S. E.
    Experimental investigation of particulate debris spreading in a pool2016Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, s. 208-219Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Termination of severe accident progression by core debris cooling in a deep pool of water under reactor vessel is considered in several designs of light water reactors. However, success of this accident mitigation strategy is contingent upon the effectiveness of heat removal by natural circulation from the debris bed. It is assumed that a porous bed will be formed in the pool in the process of core melt fragmentation and quenching. Debris bed coolability depends on its properties and system conditions. The properties of the bed, including its geometry are the outcomes of the debris bed formation process. Spreading of the debris particles in the pool by two-phase turbulent flows induced by the heat generated in the bed can affect the shape of the bed and thus influence its coolability. The goal of this work is to provide experimental data on spreading of solid particles in the pool by large-scale two-phase flow. The aim is to provide data necessary for understanding of separate effects and for development and validation of models and codes. Validated codes can be then used for prediction of debris bed formation under prototypic severe accident conditions. In PDS-P (Particulate Debris Spreading in the Pool) experiments, air injection at the bottom of the test section is employed as a means to create large-scale flow in the pool in isothermal conditions. The test section is a rectangular tank with a 2D slice geometry, it has fixed width (72 mm), adjustable length (up to 1.5 m) and allows water filling to the depth of up to 1 m. Variable pool length and depth allows studying two-phase circulating flows of different characteristic sizes and patterns. The average void fraction in the pool is determined by video recording and subsequent image processing. Particles are supplied from the top of the facility above the water surface. Results of several series of PDS-P experiments are reported in this paper. The influence of the gas flow rate, pool dimensions, particle density and size on spreading of the particles is addressed. A preliminary scaling approach is proposed and shown to provide good agreement with the experimental findings.

  • 84.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, S. E.
    Experiments and modeling of particulate debris spreading in a pool2015Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, s. 8055-8068Konferensbidrag (Refereegranskat)
    Abstract [en]

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase turbulent flows in the pool serve as a source of mechanical energy which can affect the initial geometry as well as dynamically change the shape of already formed debris bed. The main goal of this work is to provide experimental data on spreading of solid particles in the pool by large scale two-phase flow structures induced by gas injection from the bottom. These data are necessary for development and validation of predictive capabilities of computer codes allowing numerical modeling of the debris bed formation at prototypic severe accident conditions. In PDS-P experiments air injection at the bottom of the test section is employed in order to create large scale flow in the pool. The test section is constructed as a rectangular tank. It has close to 2D geometry with fixed width (72 mm), variable length (up to 1.6 m) and allows water filling depth of up to 1 m. The variable pool length and depth allows formation of the different in size and pattern two-phase circulating flows. Experimental conditions such as gas-phase flow rate and particle properties (density and size) are scaled to maintain relevancy to the prototypic accident conditions. The average void fraction in the pool is determined by video recording and image processing. Particles are supplied from the top of the facility above the water surface. In the separate-effect studies of the influence of two-phase currents on particle trajectories and bed formation, a low particle flow rate is required in order to minimize or completely exclude particle-particle interaction. Results of several series of PDS-P (Particulate Debris Spreading in the Pool) reported in this paper are analyzed analytically. The preliminary scaling approach is proposed and has good agreement with experimental findings.

  • 85.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Experimental Results on Pouring and Underwater Liquid Melt Spreading and Energetic Melt-coolant Interaction2012Ingår i: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), Kaohsiung, Taiwan, September 9-13, American Nuclear Society, 2012Konferensbidrag (Refereegranskat)
    Abstract [en]

    In a hypothetical light water reactor (LWR) core-melt accident with corium release from the reactor  vessel,  the  ultimate  containment  integrity  is  contingent  on  coolability  of  the decay-heated core debris. Pouring of melt into a pool of water located in the reactor cavity is considered in several designs of existing and new LWRs  as a part of severe accident (SA) management strategies. At certain conditions of melt release into the pool (e.g. large ratio of the  vessel  breach  size  to  the  pool  depth),  liquid  melt  can  spread  under  water  and  reach  a coolable configuration. Coolability of the melt depends on decay heat generated per unit area of the spread melt which is directly proportional to the terminal spread thickness of the melt layer. Thus a success of the debris bed coolability depends on the efficacy of the molten core materials spreading which is limited by rapid solidification  of the melt due to melt-coolant heat transfer. Among the factors which can reduce spreading effectiveness are heat and mass losses of the liquid melt due to fragmentation, cooling and solidification in the process of melt jet pouring into the pool. Previous extensive experimental and analytical works on liquid melt spreading and solidification were focused mostly on analysis of melt spreading in case of melt release through an inclined channel. Melt spreading under water as a result of a jet pouring into a pool, has not been addressed systematically. This paper summarizes first experimental results obtained in the frame of Pouring and Underwater Liquid Melt Spreading (PULiMS) research program. The work is an extension of previously reported by Kudinov et al. [1-4] studies  on debris  bed formation and  agglomeration (DEFOR-A)  phenomena.  In contrast  to DEFOR-A experiments, PULiMS exploratory tests (PULiMS-E) discussed in this work have been performed with a shallow (20 cm) water pool. Up to 78 kg of high melting temperature core melt simulant materials (eutectic mixtures of the binary oxides such as Bi 2 O 3 -WO 3  and ZrO 2 -WO 3 )  is  used  in  each  test.  Initial  melt  superheat  varied  from  70  up  to  300ºC.  In  the paper we discuss: (i) experimental observations  of the  jet  pouring into  a  shallow pool  and underwater  liquid  melt  spreading  on  a  flat  surface;  (ii)  characterization  of  solidified  melt debris;  (iii)  key  physical  processes  as  well  as  melt  material  properties  and  experimental conditions  most  influencing  the  melt  spreading  and  solidification  phenomena.  Produced experimental data can be used for validation of the models for prediction of the underwater liquid melt spreading in case of melt jet pouring in a shallow water pool.

  • 86.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Development of Scaling Approach for Prediction of Terminal Spread Thickness of Melt Poured into a Pool of Water2012Ingår i: The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), Kaohsiung, Taiwan, September 9-13, 2012, 2012Konferensbidrag (Refereegranskat)
    Abstract [en]

    Corium melt stabilization and long term cooling in a pool of water located beneath reactor vessel is adopted in several existing designs of light water reactors (LWRs) as an element in severe accident (SA) mitigation strategy. At certain conditions of melt release into the pool (e.g. large ratio of the vessel breach size to the pool depth), liquid melt can spread under water and reach a coolable configuration. Coolability of the melt is contingent on terminal spread thickness of the melt layer \delta_{sp} which defines decay heat generated per unit area of the melt surface. The thickness of the melt layer is determined by the competition between characteristic time scales of hydrodynamic melt spreading and solidification of the melt. This paper presents a modification of the scaling approach, originally proposed by Dinh et al. (2000) for prediction of the terminal melt spread thickness, to the case when liquid melt jet is poured into a pool of water and allow to spread unrestricted on a horizontal floor of the pool. Modified scaling approach takes into account mass and heat losses during to melt jet interaction with the coolant. The hydrodynamic spreading timescale is described with phenomenological approaches proposed by Huppert and Britter (1982) and Britter (1979). Proposed model is validated against PULiMS experiments (Pouring and Underwater Liquid Melt Spreading Konovalenko et al., 2012). Finally, sensitivity analysis and preliminary assessments of the uncertainties are performed for the PULiMS test conditions.

  • 87.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sköld, Per
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Controllable Generation of a Submillimeter Single Bubble in Molten Metal Using a Low-Pressure Macrosized Cavity2017Ingår i: Metallurgical and materials transactions. B, process metallurgy and materials processing science, ISSN 1073-5615, E-ISSN 1543-1916, Vol. 48, nr 2, s. 1064-1072Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    We develop a method for generation of a single gas bubble in a pool of molten metal. The method can be useful for applications and research studies where a controllable generation of a single submillimeter bubble in opaque hot liquid is required. The method resolves difficulties with bubble detachment from the orifice, wettability issues, capillary channel and orifice surfaces degradation due to contact with corrosive hot liquid, etc. The macrosized, 5- to 50-mm(3) cavity is drilled in the solid part of the pool. Flushing the cavity with gas, vacuuming it to low pressure, as well as sealing and consequent remelting cause cavity implosion due to a few orders in magnitude pressure difference between the cavity and the molten pool. We experimentally demonstrate a controllable production of single bubbles ranging from a few milliliters down to submillimeter size. The uncertainties in size and bubble release timing are estimated and compared with experimental observations for bubbles ranging within 0.16 to 4 mm in equivalent-volume sphere diameter. Our results are obtained in heavy liquid metals such as Wood's and Lead-Bismuth eutectics at 353 K to 423 K (80 A degrees C to 150 A degrees C).

  • 88.
    Kudinov, P.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Comparative testing of turbulence models of Spalart-Allmaras and Menter on the problem of transsonic flow about of single aerofoil RAE 28222004Ingår i: Bulletin of Dnipropetrovsk National University, Vol. 8, nr 1, s. 34-42Artikel i tidskrift (Refereegranskat)
    Abstract [ru]

    Проведено детальне тестування та порівняння двох найбільш поширених диференційнихмоделей турбулентності: двопараметричної моделі Ментера, та одно параметричної моделіСпаларта-Аллмараса на задачі про трансзвукове обтікання профіля RAE2822.

  • 89.
    Kudinov, Pavel
    Dnepropetrovsk State University.
    Validation of turbulence models for simulation of transonic flows with separation2005Ingår i: Transactions of XII International symposium 'Methods of discrete singularities in problems of mathematical physics' (MDOZMF-2005), 2005, s. 185-189Konferensbidrag (Refereegranskat)
  • 90.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, M.
    Approach to Prediction of Melt Debris Agglomeration Modes in a LWR Severe Accident2009Ingår i: Proceedings of ISAMM 2009: Implementation of severe accident management measures, Paul Scherrer Institute , 2009Konferensbidrag (Refereegranskat)
  • 91.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, M.
    Development and Validation of the Approach to Prediction of Mass Fraction of Agglomerated Debris2010Konferensbidrag (Refereegranskat)
  • 92.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, M.
    Development of ex-vessel debris agglomeration mode map for a LWR severe accident conditions2009Ingår i: ICONE 17: PROCEEDINGS OF THE 17TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 3, NEW YORK: AMER SOC MECHANICAL ENGINEERS , 2009, s. 37-46Konferensbidrag (Refereegranskat)
    Abstract [en]

    Ex-vessel debris bed coolability is cornerstone for severe core melt accident mitigation scheme adopted in Swedish type BWRs. Debris agglomeration can significantly affect coolability of the debris bed. The paper discusses an approach for conservative-mechanistic assessment of the debris agglomeration. In order to tackle with considerable aleatory and epistemic uncertainties which present in accident scenario and in the debris agglomeration phenomena we apply mechanistic simulation tool VAPEX code and conservative assumptions about agglomeration related phenomena and accident scenario. Results of systematic analysis arc presented in a form of agglomeration mode map. The map shows conditions for transition between different modes of debris agglomeration. Assessments of the influence of epistemic uncertainties and of numerical discretization errors on simulation result are provided. Safety implication aspects of the proposed agglomeration mode map are discussed in the present paper.

  • 93.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Davydov, M.
    Development of surrogate model for prediction of corium debris agglomeration2014Ingår i: International Congress on Advances in Nuclear Power Plants, ICAPP 2014, American Nuclear Society, 2014, s. 1146-1155Konferensbidrag (Refereegranskat)
    Abstract [en]

    Ex-vessel severe accident mitigation strategy in Nordic type Boiling Water Reactors (BWRs) imply that melt released into a deep pool of water below reactor vessel will form a coolable by natural circulation porous debris bed. However, if liquid melt is not completely fragmented and quenched when it reaches the bottom of the pool it can cause agglomeration of debris, increasing hydraulic resistance and thus worsening coolability of the bed. In the previous work we have developed and validated an approach to prediction of mass fractions of agglomerated debris using Fuel-Coolant Interaction (FCI) code VAPEX-P. This paper discusses development of a surrogate model (SM) which can predict fraction of agglomerated debris with high computational efficiency. Such model is a must for affordable sensitivity and uncertainty analysis in different accident scenarios. Details of the SM development and verification against full model are provided in the paper.

  • 94.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, M.
    Prediction of Mass Fraction of Agglomerated Debris in a LWR Severe Accident2011Konferensbidrag (Övrigt vetenskapligt)
  • 95.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Davydov, Mikhail
    Development and validation of conservative-mechanistic and best estimate approaches to quantifying mass fractions of agglomerated debris2013Ingår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 262, s. 452-461Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation, quenching, solidification and formation of a coolable by natural circulation porous debris bed in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can cause formation of agglomerated debris and "cake" regions, which affect hydraulic resistance and thus coolability of the bed. This paper discusses development and validation of conservative-mechanistic and best estimate approaches to quantifying mass fractions of agglomerated debris at given conditions of melt release from the vessel. Fuel coolant interaction (FCI) code VAPEX-P is used as a computational vehicle for modeling. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with binary oxidic simulant material melt is used for validation of developed methods. The paper discusses the influence of different inherent uncertainties in the prediction of the fraction of agglomerated debris.

  • 96.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, T.-N
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An analytical study of mechanisms that govern debris packing in a LWR severe accident2007Ingår i: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12, 2007Konferensbidrag (Refereegranskat)
    Abstract [en]

    The paper presents physical models and numerical methods developed for simulation ofadebrisbed formation duringahypotheticalsevereaccidentinaLWR. The present approach combinesamicroscale solver based on Discrete Element Method (DEM),amultiscale treatment based on "gap-tooth" scheme anda"packinglayer" model to significantly improve computational efficiency that renders simulation of the reactor-scaledebrisbed formation possible. Numerical results are presented and discussed for test problems to characterize the performance of the proposed computational engine.

  • 97.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A Computational Study of Debris Bed Formation2008Ingår i: Transaction of American Nuclear Society 2008, American Nuclear Society, 2008, s. 341-342Konferensbidrag (Refereegranskat)
  • 98.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Davydov, Mikhail
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Application of integrated deterministic-probabilistic safety analysis to assessment of severe accident management effectiveness in Nordic BWRs2016Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Konferensbidrag (Refereegranskat)
    Abstract [en]

    The goal of this work is to assess effectiveness of severe accident management strategy in Nordic type boiling water reactors (BWRs). Corium melt released into a deep pool of water below reactor vessel is expected to be fragmented to form a porous debris bed coolable by natural circulation of coolant. However, there is a risk that energetic steam explosion or non-coolable debris can threaten containment integrity. Both stochastic accident scenario (aleatory) and modeling (epistemic) uncertainties contribute to the risk assessment. Namely, the effects of melt release characteristics (jet diameter, melt composition, superheat), water pool conditions (i.e. depth and subcooling) at the time of the release, and modeling assumptions have to be quantified in a consistent manner. In order to address the uncertainty, we develop a Risk Oriented Accident Analysis framework (ROAAM+) where all stages of the accident progression are simulated using a set of models coupled through initial and boundary conditions. The analysis starts from plant damage states determined in PSA Level-1 and follows time dependent accident scenarios of core degradation, vessel failure, melt release, steam explosion and debris bed formation and coolability. In order to achieve computational efficiency sufficient for extensive sensitivity, uncertainty, and risk analysis the surrogate modeling approach is used. In the development of simplified but computationally efficient surrogate models (SM), we employ databases of solutions obtained by detailed but computationally expensive full models (FM). The process includes iterative refining of the framework, full and surrogate models in order to achieve completeness, consistency, and transparency in the review of the analysis results. In the paper we present results of the analysis aimed at quantification of uncertainty in the conditional containment failure probability. Specifically, we carry out sensitivity analysis using standalone and coupled models in order to identify the most influential scenario and modeling parameters for each sub-model. We assess the impact of the parameters on the prediction of the “load”, “capacity” and also failure probability. Then we quantify the effect of the most influential parameters on the failure probability. The results are presented using the failure domain approach and second order probability analysis, considering the uncertainty in distributions of the input parameters.

  • 99.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Yakush, Sergey
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Phung, Viet-Anh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Nam
    A framework for assessment of severe accident management effectiveness in Nordic BWR plants2014Ingår i: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Konferensbidrag (Refereegranskat)
    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.

  • 100.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Experimental investigation of debris bed agglomeration and particle size distribution using W03-ZR02 melt2015Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, s. 8046-8054Konferensbidrag (Refereegranskat)
    Abstract [en]

    Nordic BWR severe accident management strategy employs reactor cavity flooding to terminate ex-vessel accident progression. Corium melt released from the reactor pressure vessel is expected to fragment and form a porous debris bed. Success of the SAM strategy is contingent upon possibility to remove the decay heat generated in the debris bed by natural circulation of the coolant. Properties of the debris bed such as particle size, porosity and shape of the bed determine resistance for the coolant flow and thus dryout heat flux. Agglomeration of incompletely solidified debris can create additional obstacles for coolant circulation and thus reduce debris coolability margin. The goal of DEFOR (debris bed formation) experimental work is to provide data necessary for the development of analytical models and approaches for prediction of debris bed formation and agglomeration phenomena. Different corium simulant materials are used in the experiments. Liquid melt jet fragmentation and debris bed formation are considered at different conditions such as melt release (jet diameter, free fall height, etc.), melt superheat, water subcooling and water pool depth. A series of confirmatory DEFOR-A experiments has been carried out with ZrO2-WO3 simulant material. The data on particle size distribution, debris bed porosity and agglomeration is in good agreement with the previous DEFOR-S, DEFOR-A and FARO tests. On average, larger particles were obtained with ZrO2-WO3 melt than with previously used Bi2O3-WO3, size distributions for both melt simulant materials are within the ranges of size distributions observed in FARO tests. The difference between particle sizes in the tests with free falling jets was found to be insignificant. There is a tendency to form slightly larger particles only in the tests with submerged nozzles where melt is released under water with initially small jet velocity. Initial jet velocity also seems to have no visible effect on the fraction of agglomerated debris.

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