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  • 1.
    Adamsson, Carl
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    A reinterpretation of measurements in developing annular two-phase flow2011Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 241, nr 11, s. 4562-4567Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Measurements of developing films in adiabatic high pressure steam-water flow in annular geometry have been reanalyzed and compared to a linearized film-flow model. The development rate of the outer film could be determined with good accuracy in four cases. In one case it was also possible to conclude that the inner film develops faster than the outer one. When compared to the linearized model, these observations show that the deposition rate has to be almost independent of the drop concentration at the investigated conditions. Furthermore, any significant deposition by direct impaction of drops can be excluded as it would couple the development of the two films. These conclusions are quite general and do not depend on the use of any particular correlation for the deposition or entrainment rates. Finally, a rough estimate of the deposition rate was possible, confirming that deposition rates are considerably lower at high pressure steam-water flows than in air-water flows.

  • 2.
    Adamsson, Carl
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Film flow measurements for high-pressure diabatic annular flow in tubes with various axial power distributions2006Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 236, nr 23, s. 2485-2493Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Measurements of film flow rates in diabatic annular flow in tubes with various axial power distributions were carried out in the high-pressure two-phase flow loop at the Royal Institute of Technology (KTH), Sweden. The measurements were performed at conditions typical for boiling water reactors, i.e. 7 MPa pressure and total mass flux in a range from 750 to 1750 kg/m(2)s. Four different axial power distributions were used and the film mass flow was measured at 7 axial locations for each set of boundary conditions. The results show that the outlet peaked distribution gives less film than the inlet peaked one. This result is consistent with well known trends from measurements of dryout power. The measurements also show that the film flow at the onset of dryout is very small at investigated conditions in agreement with earlier studies. Finally it is shown that the present data is well predicted by two selected phenomenological models of annular flow.

  • 3.
    Adamsson, Carl
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Influence of Axial Power Distribution on Dryout: Film-Flow Models and Experiments2010Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, nr 6, s. 1495-1505Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The influence of axial power distributions on dryout occurrence in nuclear fuel assemblies has been studied extensively for several decades. Even though it is well known that axial power shapes which may significantly vary in nuclear reactors during their operation significantly change the dryout power level, this particular influence is rather difficult to capture with current correlations. In this paper it is shown that this influence can be captured using a phenomenological liquid film model coupled to a standard sub-channel code. The model has been applied to various geometries, including a round pipe, as well as 5 x 5 and 8 x 8 fuel rod assemblies, and highly accurate predictions have been obtained.

  • 4.
    Adamsson, Carl
    et al.
    Westinghouse Electric Sweden.
    Le Corre, J. M.
    Westinghouse Electric Sweden.
    Modeling and Validation of a Mechanistic Tool (MEFISTO) for the Prediction of Critical Power in BWR Fuel Assemblies2011Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 241, nr 8, s. 2843-2858Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle reaches the dryout criteria. Predicted dryout powers (including trends with flow, pressure, inlet subcooling and power distribution) and predicted dryout locations (both axial and radial) are compared to experimental results, using the entire Westinghouse SVEA-96 Optima3 dryout database, and are shown to yield excellent results.

  • 5. Almyashev, V. I.
    et al.
    Granovsky, V. S.
    Khabensky, V. B.
    Krushinov, E. V.
    Sulatsky, A. A.
    Vitol, S. A.
    Gusarov, V. V.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Barrachin, M.
    Fichot, F.
    Bottomley, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effects during corium melt in-vessel retention2016Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, s. 389-399Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 6. Alsmeyer, H
    et al.
    Albrecht, G
    Meyer, L
    Hafner, W
    Journeau, C
    Fischer, M
    Hellman, S
    Eddi, M
    Allelein, H J
    Burger, M
    Sehgal, Balraj
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Koch, M K
    Alkan, Z
    Petrov, J B
    Gaune-Escard, M
    Altstadt, E
    Bandini, G
    Ex-vessel core melt stabilization research (ECOSTAR)2005Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 235, nr 2-4, s. 271-284Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 7.
    Anghel, Ionut
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    On post-dryout heat transfer in channels with flow obstacles2014Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 270, s. 351-358Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper describes a new approach to predict post-dryout heat transfer in channels with flow obstacles. Using experimental data obtained in annular test sections at prototypical BWR conditions, the existing Saha correlation for post-dryout heat transfer has been modified to account for the presence of obstacles. The obstacle effect is taken into account in the following way: (a) the critical quality downstream of an obstacle is found; (b) an exponential function of equilibrium quality is applied to describe the decrease of heat transfer coefficient in the developing post-dryout region; (c) an additional heat transfer enhancement term is applied downstream of the obstacle. The new approach is applied to annular test sections with various flow obstacles and a significant improvement of accuracy of wall temperature prediction, as compared to reference methods, is shown.

  • 8.
    Anghel, Ionut
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Hedberg, Stellan
    KTH, Skolan för industriell teknik och management (ITM), Energiteknik, Kraft- och värmeteknologi.
    Experimental investigation of post-dryout heat transfer in annuli with flow obstacles2012Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 246, s. 82-90Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    An experimental study on post-dryout heat transfer was conducted in the High-pressure WAter Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. The objective of the experiments was to investigate the influence of flow obstacles on the post-dryout heat transfer. The investigated operational conditions include mass flux equal to 500 kg/m2 s, inlet sub-cooling 10 K and system pressure 5 and 7 MPa. The experiments were performed in annuli in which the central rod was supported with five pin spacers. Two additional types of flow obstacles were placed in the exit part of the test section: a cylinder supported on the central rod only and a typical BWR grid spacer cell. The measurements indicate that flow obstacles improve heat transfer in the boiling channel. It has been observed that the dryout power is higher when additional obstacles are present. In addition the wall temperature in post-dryout heat transfer regime is reduced due to increased turbulence and drop deposition. The present data can be used for validation of computational models of post-dryout heat transfer in channels with flow obstacles.

  • 9.
    Anglart, Henryk
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Alavyoon, Farid
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Novarini, Remi
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Study of spray cooling of a pressure vessel head of a boiling water reactor2010Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, nr 2, s. 252-257Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.

  • 10.
    Anglart, Henryk
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Andersson, Stig
    Jadrny, Reinhard
    BWR steam line and turbine model with multiple piping capability1992Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 137, s. 1-10Artikkel i tidsskrift (Fagfellevurdert)
  • 11.
    Anglart, Henryk
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Nylund, Olov
    CFD Application to prediction of void distribution in two-phase bubbly flows in rod bundles1996Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 163, s. 81-98Artikkel i tidsskrift (Fagfellevurdert)
  • 12.
    Anglart, Henryk
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Nylund, Olov
    Kurul, Necdet
    Podowski, Michael
    CFD prediction of flow and phase distribution in fuel assemblies with spacers1997Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 177, s. 215-228Artikkel i tidsskrift (Fagfellevurdert)
  • 13. Bandini, G.
    et al.
    Polidori, M.
    Gerschenfeld, A.
    Pialla, D.
    Li, S.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    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.
    Huber, K.
    Cheng, X.
    Bruzzese, C.
    Class, A. G.
    Prill, D. P.
    Papukchiev, A.
    Geffray, C.
    Macian-Juan, R.
    Maas, L.
    Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors2015Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 281, s. 22-38Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

  • 14.
    Basso, Simone
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Empirical closures for particulate debris bed spreading induced by gas-liquid flow2016Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, s. 19-25Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 15.
    Basso, Simone
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Preliminary Risk assessment of ex-vessle debris bed coolability for a Nordic BWRInngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

     

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

  • 16.
    Basso, Simone
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, S. E.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The effect of self-leveling on debris bed coolability under severe accident conditions2016Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, s. 246-259Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 17.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Kotova, S.Yu.
    Alexandrov Research Institute of Technologies (NITI).
    Kosarevsky, R.A.
    Alexandrov Research Institute of Technologies (NITI).
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Fischer, M.
    Framatome ANP GmbH.
    Corium phase equilibria based on MASCA, METCOR and CORPHAD results2008Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 238, nr 10, s. 2761-2771Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Experimental data on component partitioning between suboxidized corium melt and steel in the invessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCAprogram and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCORproject under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,. . .)–O system isreached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has twoimmiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibilitygap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,. . .)–O phase diagram investigatedwithin the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR teststhe equilibrium is established between oxidic liquid and mushy metallic part of the system. Results ofcomparison are discussed and the implications for IVR noted.

  • 18.
    Bechta, Sevostian
    et al.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymalainerf, O.
    FORTUM Nuclear Services Ltd, Espoo, Finland.
    VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere2009Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, nr 6, s. 1103-1112Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  • 19.
    Bechta, Sevostian
    et al.
    Scientific Research Technological Institute (NITI), Russian Federation.
    Khabensky, V. B.
    Vitol, S. A.
    Krushinov, E. V.
    Lopukh, D. B.
    Petrov, Yu.B.
    Petchenkov, A.Yu.
    Kulagin, I. V.
    Granovsky, V. S.
    Kovtunova, S. V.
    Martinov, V. V.
    Gusarov, V. V.
    Experimental studies of oxidic molten corium-vessel steel interaction2001Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 210, nr 1-3, s. 193-224Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  • 20.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Fieg, G.
    Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Tuomisto, H.
    Fortum Engineering Ltd..
    Corrosion of vessel steel during its interaction with molten corium: Part 1. Experimental2006Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 236, nr 17, s. 1810-1829Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheresduring an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities andoxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimeningot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.

  • 21.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Fieg, G.
    Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Tuomisto, H.
    Fortum Engineering Ltd..
    Corrosion of vessel steel during its interaction with molten corium: Part 2. Model development2006Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 236, nr 13, s. 1362-1370Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments havebeen conducted on “Rasplav-2” test facility and followed up with physico-chemical and metallographic analyses of melt samples and coriumspecimeningots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere abovethe melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate orcorrosion depth of vessel steel in conditions simulated by the experiments.

  • 22.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Kotova, S.Yu.
    Alexandrov Research Institute of Technologies (NITI).
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almyashev, V.I.
    Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences.
    Ducros, G.
    CEA, DEN, Cadarache.
    Journeau, C.
    CEA, DEN, Cadarache.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Clément, B.
    Institut de Radioprotection et Sûreté Nucléaire.
    Herranz, L.
    CIEMAT.
    Guentay, S.
    Paul Scherrer Institut (PSI).
    Trambauer, K.
    GemResearch Swisslab (GRS).
    Auvinen, A.
    Technical Research Centre of Finland (VTT).
    Bezlepkin, V.V.
    SPbAEP.
    Influence of corium oxidation on fission product release from molten pool2010Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, nr 5, s. 1229-1241Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Qualitative and quantitative determination of the release of low-volatile fission products and core materialsfrom molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. Theexperiments carried out in a cold crucible with induction heating and RASPLAV test facility are described.The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidationkinetics, critical influence of melt surface temperature and oxidation index on the fission productrelease rate, aerosol particle composition and size distribution. The relevance of measured high releaseof Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimentaldata with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions fromIVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations areproposed following the major observations and discussions.

  • 23.
    Bechta, Sevostian
    et al.
    Sci. Res. Technol. Institute (NITI), Russian Federation.
    Vitol, S. A.
    Krushinov, E. V.
    Granovsky, V. S.
    Sulatsky, A. A.
    Khabensky, V. B.
    Lopukh, D. B.
    Petrov, Y. B.
    Pechenkov, A. Y.
    Water boiling on the corium melt surface under VVER severe accident conditions2000Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 195, nr 1, s. 45-56Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the `Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO2+x-16% ZRO2-15% Fe2O3-6% Cr2O3-3% Ni2O3. The melt surface temperature ranged within 1920-1970 K.

  • 24.
    Bergagio, Mattia
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi. Warsaw University of Technology, Poland.
    Experimental investigation of mixing of non-isothermal water streams at BWR operating conditions2017Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 317, s. 158-176Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this experimental investigation, wall surface temperatures have been measured during mixing of three water streams in the annular gap between two coaxial stainless-steel tubes. The inner tube, with an outer diameter of 35 mm and a thickness of 5 mm, holds six K-type, ungrounded thermocouples with a diameter of 0.5 mm, which measured surface temperatures with a sampling rate of either 100 Hz or 1000 Hz. The tube was rotated from 0 to 360° and moved in a range of 387 mm in the axial direction to allow measurements of surface temperatures in the whole mixing region. The outer tube has an inner diameter of 80 mm and a thickness of 10 mm to withstand a water pressure of 9 MPa. A water stream at a temperature of either 333 K or 423 K and a Reynolds number between 1657 and 8410 rose vertically in the annular gap and mixed with two water streams at a temperature of 549 K and a Reynolds number between 3.56E5 and 7.11E5. These two water streams entered the annulus radially on the same axial level, 180° apart. Water pressure was kept at 7.2 MPa. Temperature recordings were performed at five axial and eight azimuthal locations, for each set of boundary conditions. Each recording lasted 120 s to provide reliable data on the variance, intermittency and frequency of the surface temperature time series at hand. Thorough calculations indicate that the uncertainty in the measured temperature is of 1.58 K. The mixing region extends up to 0.2 m downward of the hot inlets. In most cases, measurements indicate non-uniform mixing in the azimuthal direction, because of asymmetries in either geometry or mass flow rates at the hot inlets. Due to the measurement accuracy and a relatively simple geometry, an experimental database has been obtained for validation of computational methods to predict thermal mixing and fatigue. Furthermore, these data can provide new insight into turbulent mixing at BWR operating conditions and, more generally, into mixing coupled to the dynamics, also termed level-2 mixing.

  • 25. Bottomley, D.
    et al.
    Stuckert, J.
    Hofmann, P.
    Tocheny, L.
    Hugon, M.
    Journeau, C.
    Clement, B.
    Weber, S.
    Guentay, S.
    Hozer, Z.
    Herranz, L.
    Schumm, A.
    Oriolo, F.
    Altstadt, E.
    Krause, M.
    Fischer, M.
    Khabensky, V. B.
    Bechta, Sevostian V.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Veshchunov, M. S.
    Palagin, A. V.
    Kiselev, A. E.
    Nalivaev, V. I.
    Goryachev, A. V.
    Zhdanov, V.
    Baklanov, V.
    Severe accident research in the core degradation area: An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center2012Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 252, s. 226-241Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 26. Bubelis, E.
    et al.
    Tosello, A.
    Pfrang, W.
    Schikorr, M.
    Mikityuk, K.
    Panadero, A. -L
    Martorell, S.
    Ordóñez, J.
    Seubert, A.
    Lerchl, G.
    Stempniewicz, M.
    Alcaro, F.
    De Geus, E.
    Delmaere, T.
    Poumerouly, S.
    Wallenius, Janne
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions2017Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 320, s. 325-345Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.

  • 27.
    Caraghiaur, Diana
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Adamsson, Carl
    Paul Scherrer Institute.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    A model for inertial drop deposition suitable to predict obstacle effect2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 260, s. 121-133Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The drop deposition increase due to flow obstruction is usually quantified solely by empirical coefficients. In this work we propose a new way to calculate the drop deposition with the capability to predict the obstacle effect. The model is based on the drop volume fraction, slip ratio and turbulence quantities of the continuous phase obtained from the two-fluid calculations. Additional relations are needed to calculate the fluctuating velocities of the drop phase. These relations are based on the fluid integral time scales. A number of relations are tested, which include the effect of drop inertia and drift parameter. The new model is tested in a number of flow combinations, including air-water and helium-water of 1.5 bar and steam-water at 70 bar pressure, for low and high drop concentration. The high concentration flow shows that further studies are needed to include drop size increase due to coalescence and reduction of velocity fluctuation due to drop collisions. The new model is tested for pipe flow containing an obstacle of steam-water flows of 5, 10 and 15 bar pressure. The new model shows the capability to qualify the obstacle effect. Further improvements are needed to increase the quantitative capability.

  • 28.
    Caraghiaur, Diana
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Drop deposition in annular two-phase flow calculated with Lagrangian Particle Tracking2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 265, s. 856-866Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Lagrangian Particle Tracking is tested for its capability to predict deposition rates in pipes and pipes with obstacle. The drop size is one of the input parameters, which defines in its major part the deposition process. A new correlation is proposed to estimate the drop size, following a systematic analysis of the experimental drop sizes in annular twophase flow. The Lagrangian Particle Tracking model showed good capability of prediction in the cases where the drop size is known; however, when the drop size is estimated the inaccuracy in calculated deposition rate is high. If the drop size is known at the inlet of the channel, Lagrangian Particle Tracking shows good capability of predicting the deposition increase downstream of the obstacle for steam-water flows of 5, 10 and 15 bar pressure.

  • 29.
    Caraghiaur, Diana
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Frid, Wiktor
    Experimental investigation of turbulent flow through spacer grids in fuel rod bundles2009Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, nr 10, s. 2013-2021Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper contains experimental data of pressure, velocity and turbulence intensity in a 24‐rod fuel bundle withspacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressuresensingrod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuatingcomponent upstream and downstream of the spacer grid in subchannels with different blockage ratios. Themeasurements show a changing pattern in function of radial position in the cross‐section of the fuel bundle. Forsubchannels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer andthen gradually decays. In inner subchannels, however, the turbulence intensity downstream of the spacer decreasesbelow its upstream value and then gradually increases until it reaches the maximum value at approximately twospacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulenceintensity enhancement, do not depend exclusively on the local geometry details, but also on the location in thecross‐section of the rod bundle.

  • 30. Cheng, X.
    et al.
    Batta, A.
    Bandini, G.
    Roelofs, F.
    Van Tichelen, K.
    Gerschenfeld, A.
    Prasser, M.
    Papukchiev, A.
    Hampel, U.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems2015Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, s. 2-12Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

  • 31.
    Concilio Hansson, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Manickam, Louis
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A study of the effect of binary oxide materials in a single droplet vapor explosion2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 264, s. 168-175Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 32. Dombrovsky, Leonid A.
    et al.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effect of thermal radiation on the solidification dynamics of metal oxide melt droplets2008Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 238, nr 6, s. 1421-1429Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 33.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kapulla, R.
    Paranjape, S.
    Paladino, D.
    Laine, J.
    Puustinen, M.
    Räsänen, A.
    Pyy, L.
    Kotro, E.
    Pool stratification and mixing during a steam injection through spargers: analysis of the PPOOLEX and PANDA experiments2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 337, s. 300-316Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors to condense steam in large water pools. A steam injection induces heat, momentum and mass sources that depend on the steam injection conditions and can result in thermal stratification or mixing of the pool. Thermal stratification reduces the steam condensation capacity of the pool, increases the pool surface temperature and thus the containment pressure. Development of models with predictive capabilities requires the understanding of basic phenomena that govern the behavior of the complex multi-scale system. The goals of this work are (i) to analyze and interpret the experiments on steam injection into a pool through spargers performed in the large-scale facilities of PPOOLEX and PANDA, and (ii) to discuss possible modelling approaches for the observed phenomena. A scaling approach was developed to address the most important physical phenomena and regimes relevant to prototypic plant conditions. The focus of the tests was on the low steam mass flux and oscillatory bubble condensation regimes, which are expected during a long-term steam injection transient, e.g. in the case of a Station Black Out (SBO). Exploratory tests were also done for chugging and stable jet conditions. The results showed a similar behavior in PPOOLEX and PANDA in terms of jet induced by steam condensation, pool stratification, and development of hot layer and erosion of the cold one. A correlation using the Richardson number is proposed to model the erosion rate of the cold layer as a function of the pool dimensions and steam injection conditions.

  • 34.
    Galushin, Sergey
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova Univ Ctr, SE-10691 Stockholm, Sweden.
    Sensitivity analysis of debris properties in lower plenum of a Nordic BWR2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, s. 374-382Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 35. Granovsky, V. S.
    et al.
    Khabensky, V. B.
    Krushinova, E. V.
    Vitol, S. A.
    Sulatsky, A. A.
    Almjashev, V. I.
    Bechta, Sevostian V.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gusarov, V. V.
    Barrachin, M.
    Bottomle, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention2014Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 278, s. 310-316Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO2+x-ZrO2-FeOy corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 degrees C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR).

  • 36.
    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 codes2015Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, s. 144-153Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 37.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation and application of the MEWA code to analysis of debris bed coolability2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, s. 22-37Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper was first aimed at validating the MEWA code against experiments for two-phase flow and dryout in particulate beds, and then investigating the coolability of ex-vessel debris beds with cylindrical, conical and truncated conical shapes assumed to form under severe accident scenarios of a boiling water reactor. The validation was mainly performed against the POMECO-FL and POMECO-HT experiments carried out at KTH for investigating frictional laws and coolability limit (dryout) of particulate beds, respectively. The comparison of the experimental and numerical results shows that the MEWA code is capable of predicting both the pressure drop of two-phase flow through porous media and the dryout condition of various stratified beds. While the coolability of a one-dimensional homogeneous debris bed is bounded by counter-current flow limit (CCFL), the coolability of a heap-like debris bed can be improved due to lateral ingression of coolant in a multi-dimensional geometry. The simulations showed that the dryout power density of a prototypical debris bed was roughly inversely proportional to the bed's height regardless of the bed's shape. The impacts of a debris bed's features on coolability are manifested in three aspects: multidimensionality and contour surface area of the bed, as well as the uniformity of its shape. The contour surface area is defined as the interface between debris bed and water pool, and its effective value depends on the surface orientation that determines the amount of water ingress and vapor escape. The perfect uniformity in bed's shape as cylindrical bed results in even distributions of temperature and void fraction. The dryout power density was also predicted to be strongly correlated to the uniformity of bed's shape. The MEWA simulation also predicted that coolability was improved by an downcomer embedded in the center of debris bed. The efficiency of such enhancement was largely determined by the downcomer's length, whose optimal value was obtained in simulation.

  • 38.
    Jaromin, Maria
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    A numerical study of heat transfer to supercritical water flowing upward in vertical tubes under normal and deteriorated conditions2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 264, s. 61-70Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    A numerical study of heat transfer to supercritical water in vertical tubes is carried out using the ANSYS-CFX code and employing the k-omega SST turbulence model. The numerical results on wall temperature distributions under normal and deteriorated heat transfer conditions are compared with experimental data and a good agreement is obtained. The onset of deterioration is captured for both low-flow and high-flow conditions. Sensitivity of numerical results to operational conditions and the turbulent Prandtl number (Pr-t) is investigated. The grid independent solution is obtained when y(+) is less than 1 and the cell aspect ratio is less than 2000. It is concluded that the turbulent Prandtl number has a quite significant influence on the calculated wall temperature and the best agreement with experimental data is obtained when Pr-t is close to 0.9.

  • 39.
    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 experiment2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikkel i tidsskrift (Fagfellevurdert)
  • 40.
    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-3DInngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikkel i tidsskrift (Fagfellevurdert)
    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.

  • 41.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Seismic sloshing effects in lead-cooled fast reactors2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, s. 99-110Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 42.
    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.
    Seismic sloshing effects in lead-cooled fast reactors2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, s. 99-110Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 43.
    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 core2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 328, s. 255-265Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 44.
    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 facility2009Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, nr 9, s. 1653-1659Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 45.
    Karkoszka, Krzysztof
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Multidimensional effects in laminar filmwise condensation of vapor in binary and ternary mixtures with non-condensable gases2008Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 238, nr 6, s. 1373-1381Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper is dealing with the multidimensional modelling of gravity driven water vapour free convection condensation from binary and ternary mixtures of condensable and noncondensable gases. In the case of ternary mixtures, a special attention is paid to the influence of the light gas on the transport phenomena in the gaseous phase. Two solution methods have been applied: an analytical solution employing the boundary layer similarity approximation and a numerical solution of multi-fluid, multi-component formulation of the conservation equations. It has been demonstrated that the two methods are equivalent when applied to binary mixtures in simple geometries. However, to capture the spatial effects and the ternary mixture phenomena the latter method must be used.

  • 46. Khabensky, V. B.
    et al.
    Granovsky, V. S.
    Almjashev, V. I.
    Vitol, S. A.
    Krushinov, E. V.
    Kotova, S. J.
    Sulatsky, A. A.
    Gusarov, V. V.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Barrachin, M.
    Bottomley, D.
    Fischer, M.
    Hellmann, S.
    Piluso, P.
    Miassoedov, A.
    Tromm, W.
    Effect of temperature gradient on chemical element partitioning in corium pool during in-vessel retention2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, s. 82-91Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The paper presents some results of the ISTC (International Science and Technology Center)-financed project ‘Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel’ (METCOR). In the METCOR experiments the metallic phase of a two-liquid system was produced by the interaction between hot suboxidized corium and cooled VVER vessel steel, with the steel being corroded. Models of corrosion mechanisms in the considered conditions are used to systematize data on the limiting temperature of corrosion/(dissolution) of the vessel steel. A considerable influence of thermal gradient conditions is shown, which has to be taken into account in the analysis of molten pool behaviour. 

  • 47.
    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 pool2016Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, s. 208-219Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 48.
    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 debris2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 262, s. 452-461Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 49.
    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.
    Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials2017Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 314, s. 182-197Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

  • 50.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Agglomeration and size distribution of debris in DEFOR-A experiments with Bi2O3-WO3 corium simulant melt2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 263, s. 284-295Artikkel i tidsskrift (Fagfellevurdert)
    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.

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