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  • 1. Bandini, G.
    et al.
    Bubelis, E.
    Schikorr, M.
    Stempnievicz, M. H.
    Lázaro, A.
    Tucek, K.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Mansani, L.
    Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The conceptual design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is under development within the LEADER project to meet the safety objectives of Gen IV nuclear energy systems. This paper presents the main results of the safety analysis for beyond design basis conditions, namely design extension conditions (DEC), which include the failure of prevention and mitigation systems, like the reactor scram in the so called unprotected transients. The main objective of this analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the ALFRED reactor. Several computer codes: SIM LFR, RELAP5, CATHARE, SPECTRA and TRACE are applied to evaluate the consequences of representative unprotected accident scenarios such as Loss of Flow, Loss of Heat Sink and Reactivity initiated accidents. Additionally, the consequences of steam generator tube rupture and partial sub assembly flow blockage events are assessed by means of appropriate fluid dynamic codes. 

  • 2. Bandini, G.
    et al.
    Polidori, M.
    Gerschenfeld, A.
    Pialla, D.
    Li, S.
    Ma, Weimin
    KTH, 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.

  • 3.
    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.

  • 4.
    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 Codes2014Konferansepaper (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, (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.

  • 5.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors2018Doktoravhandling, med artikler (Annet vitenskapelig)
    Abstract [en]

    Carbon-free nuclear power production can help to relieve the increasing energy demand in the world. New generation reactors with improved safety, sustainability, reliability, economy and security are being developed. Lead-cooled fast reactor (LFR) is one of them.

    Lead-cooled reactors have many advantages related to the use of liquid metal coolants and pool-type primary system designs. Still, several technological challenges are to be overcome before their successful commercial deployment. The decisions on new designs and licensing of LFRs require use of code simulations, due to lack of operational experience and the fact that full scale experimental investigations are impractical. Code development and validation is necessary in order to reduce uncertainty in predictions for risk assessment and decision support. In order to make the decisions robust, i.e. not sensitive to remaining uncertainty, the process of collecting necessary evidences should iteratively converge. Extensive sensitivity and uncertainty analysis is necessary in order to make the process user-independent.

    The pool-type design introduces 3D phenomena that require application of advanced modeling tools such as Computational Fluid Dynamics (CFD). The focus of this thesis is on the development and application of CFD modeling and general code validation methodology to the following issues of safety importance for LFR: (i) pool thermal stratification and mixing, (ii) steam generator tube leakage, (iii) seismic sloshing, and (iv) solidification.

    Thermal mixing and stratification in a pool-type reactor can affect natural circulation stability and thus decay heat removal capability and pose thermal stresses on the reactor structures. Standalone System Thermal-hydraulics (STH) and CFD codes are either inadequate or too computationally expensive for analysis of such phenomena in prototypic conditions. In this work an approach to coupling of STH and CFD is proposed and implemented. For validation of standalone and coupled codes TALL-3D facility (lead-bismuth eutectic (LBE) loop) and test matrix was designed featuring significant two-way thermal-hydraulic feedbacks between the pool-type test section and the rest of the loop.

    The possibility of core voiding due to bubble transport in case of steam generator tube leakage/rupture (SGTL/R) is a serious safety concern that can be a showstopper for LFR licensing. Voiding of the core increases risks for reactivity insertion accident (RIA) and/or local fuel damage due to deteriorated heat transfer. Bubble transport analysis for the ELSY (European Lead-cooled SYstem) reactor design demonstrated the effect of uncertainty in drag correlation, bubble size distributions and leak location on the core and primary system voiding. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.

    The effect of seismic isolation (SI) system on the sloshing of the heavy metal coolant and respective mechanical loads on the vessel structures is studied for design basis and beyond design basis earthquakes. It was found that SI shifts the frequencies closer to resonance conditions for sloshing, thus increasing the loads due to slamming of coolant waves onto the vessel internal structures including reactor lid. Possible mitigation measures, such as partitioning baffles have been proposed and their effectiveness was analyzed. 

    Coolant solidification is another serious safety issue in liquid metal cooled reactors that can cause partial or even full blockage of the coolant flow path and thermo-mechanical loads on the primary system structures. Validation of melting/solidification models implemented in CFD is a necessary pre-requisite for its application to risk analysis. In this work, the parametric analysis in support of the design of a solidification test section (STS) in TALL3D facility is carried out. The goals and requirements for the validation experiment and how they are satisfied in the design are discussed in the thesis.

  • 6. 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.2011Konferansepaper (Fagfellevurdert)
  • 7.
    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)
  • 8.
    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 experiment2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 341, s. 306-325Artikkel i tidsskrift (Fagfellevurdert)
    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).

  • 9. 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 Reactors2011Inngår i:  , 2011Konferansepaper (Fagfellevurdert)
  • 10.
    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 codes2012Inngår i: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Konferansepaper (Fagfellevurdert)
  • 11.
    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 Codes2012Inngår i: Proceedings of The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 2012Konferansepaper (Fagfellevurdert)
  • 12.
    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.

  • 13.
    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-3D2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 339, s. 21-38Artikkel i tidsskrift (Fagfellevurdert)
    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. 

  • 14.
    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-hydraulics2013Inngår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 15.
    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 experiment2012Konferansepaper (Fagfellevurdert)
  • 16.
    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 Experiment2014Konferansepaper (Fagfellevurdert)
    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.

  • 17.
    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 reactor2015Inngår i: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 190, nr 1, s. 1-10Artikkel i tidsskrift (Fagfellevurdert)
    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.

  • 18.
    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 reactor2014Inngår i: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Konferansepaper (Fagfellevurdert)
    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.

  • 19.
    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.

  • 20.
    Kööp, Kaspar
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    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 for identification of natural circulation instabilties in TALL-3D facility2017Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, Vol. 314, s. 110-120Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    TALL-3D facility is a lead-bismuth eutectic (LBE) thermal-hydraulic loop designed to provide experimental data on thermal-hydraulics phenomena for validation of stand-alone and coupled System Thermal Hydraulics (STH) and Computational Fluid Dynamics (CFD) codes. Pre-test analysis is crucial for proper choice of experimental conditions at which the experimental data would be most useful for code validation and benchmarking. The goal of this work is to identify these conditions at which the experiment is challenging for the STH codes yet minimizes the 3D-effects from the test section on the loop dynamics. The analysis is focused on the identification of limit cycle flow oscillations in the TALL-3D facility main heater leg using a global optimum search tool GA-NPO to find a general region in the parameter space where oscillatory behavior is expected. As a second step a grid study is conducted outlining the boundaries between different stability modes. Phenomena, simulation results and methodology for selection of the test parameters are discussed in detail and recommendations for experiments are provided.

  • 21.
    Mickus, Ignas
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    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.
    Lappalainen, J.
    Development of tall-3d test matrix for APROS code validation2015Inngår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, s. 4562-4575Konferansepaper (Fagfellevurdert)
    Abstract [en]

    APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.

  • 22.
    Mickus, Ignas
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Vorobyev, Yuri
    Moscow Power Engineering Institute.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    An Approach to Physics Based Surrogate Model Development for Application with IDPSA2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 23.
    Moreau, V.
    et al.
    CRS4, Sci & Technol Pk Polaris Piscina Manna, I-09010 Pula, Italy..
    Profir, M.
    CRS4, Sci & Technol Pk Polaris Piscina Manna, I-09010 Pula, Italy..
    Alemberti, A.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Frignani, M.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Merli, F.
    Ansaldo Nucl SpA, Corso Perrone,25, I-16161 Genoa, Italy..
    Belka, M.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Frybort, O.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Melichar, T.
    CVR, Hlavni 130, Husinec Rez 25068, Czech Republic..
    Tarantino, M.
    ENEA FSN ING, I-40033 Camugnano, BO, Italy..
    Franke, S.
    HZDR, Bautzner Landstr 400, D-01328 Dresden, Germany..
    Eckert, S.
    HZDR, Bautzner Landstr 400, D-01328 Dresden, Germany..
    Class, A.
    KIT, Kaiserstr 12, D-76131 Karlsruhe, Germany..
    Yanez, J.
    KIT, Kaiserstr 12, D-76131 Karlsruhe, Germany..
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Roelofs, F.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Zwijsen, K.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Visser, D. C.
    NRG, Westerduinweg 3, NL-1755 LE Petten, Netherlands..
    Badillo, A.
    PSI, CH-5232 Villigen, Switzerland..
    Niceno, B.
    PSI, CH-5232 Villigen, Switzerland..
    Martelli, D.
    UNIPI, DICI, Largo Lucio Lazzarino,2, I-56122 Pisa, Italy..
    Pool CFD modelling: lessons from the SESAME project2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 355, artikkel-id UNSP 110343Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.

  • 24.
    Moreau, V.
    et al.
    CRS4, Science and Technology Park Polaris – Piscina Manna, Pula, 09010, Italy.
    Profir, M.
    CRS4, Science and Technology Park Polaris – Piscina Manna, Pula, 09010, Italy.
    Alemberti, A.
    CRS4, Science and Technology Park Polaris – Piscina Manna, Pula, 09010, Italy.
    Frignani, M.
    Ansaldo Nucleare SpA, Corso Perrone, 25, Genova, 16161, Italy.
    Merli, F.
    Ansaldo Nucleare SpA, Corso Perrone, 25, Genova, 16161, Italy.
    Belka, M.
    CVR, Hlavni 130, Husinec-Rez, 250 68, Czech Republic.
    Frybort, O.
    CVR, Hlavni 130, Husinec-Rez, 250 68, Czech Republic.
    Melichar, T.
    CVR, Hlavni 130, Husinec-Rez, 250 68, Czech Republic.
    Tarantino, M.
    ENEA FSN-ING, Brasimone R.C., Camugnano (Bo), 40033, Italy.
    Franke, S.
    HZDR, Bautzner Landstrasse 400, Dresden, 01328, Germany.
    Eckert, S.
    HZDR, Bautzner Landstrasse 400, Dresden, 01328, Germany.
    Class, A.
    KIT, Kaiserstr. 12, Karlsruhe, 76131, Germany.
    Yanez, J.
    KIT, Kaiserstr. 12, Karlsruhe, 76131, Germany.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Roelofs, F.
    NRG, Westerduinweg 3, LE Petten, 1755, Netherlands.
    Zwijsen, K.
    NRG, Westerduinweg 3, LE Petten, 1755, Netherlands.
    Visser, D. C.
    NRG, Westerduinweg 3, LE Petten, 1755, Netherlands.
    Badillo, A.
    PSI, Villigen, 5232, Switzerland.
    Niceno, B.
    PSI, Villigen, 5232, Switzerland.
    Martelli, D.
    UNIPI DICI Largo Lucio Lazzarino, 2 Pisa, 56122, Italy.
    Pool CFD modelling: Lessons from the sesame project2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 355, artikkel-id 110343Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from. CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.

  • 25. Papukchiev, Angel
    et al.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Geffray, Clotaire
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Macian, Rafael
    Lerchl, Georg
    Prediction of Complex Thermal-Hydraulic Phenomena Supplemented by Uncertainty Analysis with Advanced Multiscale Approaches for the TALL-3D T01 Experiment2014Inngår i: Proceedings of the 12th International Probabilistic Safety Assessment and Management Conference (PSAM 12), Techno-Info Comprehensive Solutions (TICS) , 2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The thermal-hydraulic (TH) system code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. No tests have been performed up to now. GRS carried out pre-test simulations with ATHLET - ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET-ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared. Moreover, this work is supplemented by uncertainty and sensitivity analysis for the T01 experiment, carried out at the Technische Universitaet Muenchen.

  • 26. Papukchiev, Angel
    et al.
    Jeltsov, Marti
    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.
    Lerchl, Georg
    Comparison of different coupling CFD-STH approaches for pre-test analysis of a TALL-3D experiment2015Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, s. 135-143Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The system thermal-hydraulic (STH) code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. GRS carried out pre-test simulations with ATHLET-ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET-ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared.

  • 27. Papukchiev, Angel
    et al.
    Lerchl, Georg
    Geffray, C.
    Jeltsov, Marti
    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.
    Coupled 1D-3D Thermal-Hydraulic Simulations of a Liquid Metal Experiment Supplemented by Uncertainty and Sensitivity Analysis2014Inngår i: OECD/NEA & IAEA Workshop: Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation, 2014Konferansepaper (Fagfellevurdert)
  • 28. Vorobyev, Yu.B.,
    et al.
    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.
    Nhat, T.V.K.
    Application of information technologies (genetic algorithms, neural networks, parallel calculations) in safety analysis of Nuclear Power Plants2014Inngår i: Proceedings of the Institute for System Programming, ISSN 2220-6426, Vol. 26, nr 2, s. 137-158Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper investigates important issues in three types of safety assessment methodologies commonly applied for Nuclear Power Plants (NPP). These methodologies are i) dynamic probabilistic safety assessment (DPSA) where application of genetic algorithm (GA) is shown to improve the efficiency of the analysis, ii) deterministic safety assessment (DSA) with meta model representation of the system using pre-performed computational fluid dynamics (CFD) code and iii) vulnerability search (e.g. identification of accident scenarios in an NPP) with application of neural network (NN). The use of advanced computational tools and methods such as genetic algorithms, neural networks and parallel computations improve the efficiency of safety analysis. To achieve the best effect, these advanced technologies are to be integrated with existing classical methods of safety analysis of the NPP.

1 - 28 of 28
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