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  • 1.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    Kudinov, Pavel
    Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression PoolManuskript (preprint) (Annet vitenskapelig)
    Abstract [en]

    The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.

  • 2.
    Gallego-Marcos, Ignacio
    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.
    Thermal stratification and mixing in a Nordic BWR pressure suppression pool2019Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 132, s. 442-450Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.

  • 3.
    Galushin, Sergey
    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.
    Implementation of Probabilistic Framework of Risk Analysis Framework for Assessment of Severe Accident Management Effectiveness in Nordic BWRInngår i: Annals of Nuclear EnergyArtikkel i tidsskrift (Fagfellevurdert)
  • 4.
    Galushin, Sergey
    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.
    Risk analysis framework for severe accident mitigation strategy in nordic BWR: An approach to communication and decision making2017Inngår i: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, American Nuclear Society , 2017, s. 587-594Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 5.
    Galushin, Sergey
    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.
    Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code2019Konferansepaper (Fagfellevurdert)
  • 6.
    Galushin, Sergey
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR2019Konferansepaper (Fagfellevurdert)
  • 7.
    Galushin, Sergey
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Ranlöf, L.
    Bäckström, O.
    Adolfsson, Y.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Marklund, A. R.
    Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR2018Inngår i: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Konferansepaper (Fagfellevurdert)
    Abstract [en]

    A comprehensive and robust assessment of severe accident management effectiveness in preventing unacceptable releases is a challenge for a today’s real life PSA. This is mainly due to the fact that major uncertainty is determined by the physical phenomena and timing of the events. The static PSA is built on choosing scenario parameters to describe the accident progression sequence and typically uses a limited number of simulations in the underlying deterministic analysis. Risk Oriented Accident Analysis Methodology framework (ROAAM+) is being developed in order to enable consistent and comprehensive treatment of both epistemic and aleatory uncertainties. The framework is based on a set of deterministic models that describe different stages of the accident progression. The results are presented in terms of distributions of conditional containment failure probabilities for given combinations of the scenario parameters. This information is used for enhanced modeling in the PSA-L2. Specifically, it includes improved definitions of the sequences determined by the physical phenomena rather than stochastic failures of the equipment, improved knowledge of timing in sequences and estimation of probabilities determined by the uncertainties in the phenomena. In this work we present an example of application of the dynamic approach in a large scale PSA model and show that the integration of the ROAAM+ results and the PSA model can potentially lead to a considerable change in PSA Level 2 analysis results. 

  • 8.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs2016Inngår i: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, s. 311-327Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 9.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR2017Inngår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 10.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Risk of containment failure due to ex-vessel steam explosion for Nordic BWRs2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, s. 4032-4038Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In case of a severe accident in a Light Water Reactor (LWR) degraded core relocates into the lower head of the reactor pressure vessel. Under thermal and mechanical loads from the core debris the vessel can fail releasing hot debris into the containment. In some designs of LWRs the severe accident mitigation strategy aims to prevent early containment failure by providing a pool of water below the reactor vessel. The melt is expected to form a coolable debris configuration preventing or delaying release of radioactive materials to the environment. One of the risk factors associated with melt-water interaction is containment failure due to ex-vessel steam explosion. Energetics of the steam explosion is contingent upon characteristics of melt release, pool and containment geometry. A general purpose full and surrogate models for estimation of the steam explosion loads in various conditions prototypic to boiling and pressurized water reactors have been proposed. In this paper, we rely on our recent results in model validation to develop a new surrogate model for the estimation of the steam explosion loads in LWRs using less conservative assumptions. We sample model output using Risk Oriented Accident Analysis Methodology code (ROAAM+) and provide estimates for the risk of containment failure for Nordic BWR given different accident scenarios. We plot Failure Domain maps and discuss implication of the steam explosion for different designs (fragility levels) and severe accident management strategies (pool depths). Importantly, we analyze the effect of the reduced model conservatism on the results of the risk analysis and discuss its implications to the decision making.

  • 11.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of a full model for the analysis of ex-vessel steam explosion in LWRs2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, s. 4568-4574Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In a Light Water Reactor (LWR) severe accident, the reactor core can be melted and released from the reactor vessel at ~3000K. In most of reactor designs ex-vessel severe accident mitigation strategy employs a pool of water underneath the reactor vessel. If water pool is deep enough, the melt is expected to be fragmented and quenched and form a coolable debris bed preventing further accident progression. However, there is a possibility that upon contact with volatile coolant thermal energy stored in the hot melt will be converted into mechanical energy of rapidly expanding steam in the process of so called “steam explosion”. Energetics of the steam explosion is contingent upon conditions of melt release, pool characteristics and containment geometry. Containment failure due to the ex-vessel steam explosion can be a factor of risk for the “wet cavity” strategy if fragility limits are close to the expected loads. In order to assess the risk, we develop so called full model (based on TEXAS-V code) for the estimation of the steam explosion loads. To ensure model applicability to a wide range of LWR designs, a number of modifications have been introduced in comparison to previous works. A large database of Full Model solutions is used then for the development of a Surrogate Model based on the Artificial Neural Networks (ANN) to enable extensive sensitivity analysis and uncertainty quantification. The uncertainty in the SM approximation of the FM is considered explicitly in the assessment of failure probability. In this work, we demonstrate an approach to the validation of the Full Model against previous steam explosion experiments using a statistical approach in which a joint distribution of the experimental data is compared to a database of explosion distributions obtained using the full model.

  • 12.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. Royal Inst Technol, Div Nucl Engn, Roslagstullsbacken 21, SE-10691 Stockholm, Sweden..
    Papukchiev, A.
    Liu, C.
    Geffray, C.
    Polidori, M.
    Koop, Björn
    KTH, Skolan för teknikvetenskap (SCI), Tillämpad fysik, Nanostrukturfysik.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    TALL-3D open and blind benchmark on natural circulation instability2020Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 358, artikkel-id 110386Artikkel i tidsskrift (Fagfellevurdert)
  • 13.
    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)
  • 14.
    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).

  • 15.
    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.

  • 16.
    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. 

  • 17.
    Konovalenko, Alexander
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sköld, Per
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Controllable Generation of a Submillimeter Single Bubble in Molten Metal Using a Low-Pressure Macrosized Cavity2017Inngår i: Metallurgical and materials transactions. B, process metallurgy and materials processing science, ISSN 1073-5615, E-ISSN 1543-1916, Vol. 48, nr 2, s. 1064-1072Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 18.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Davydov, Mikhail
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Application of integrated deterministic-probabilistic safety analysis to assessment of severe accident management effectiveness in Nordic BWRs2016Inngår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 19.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yakush, Sergey E.
    I nstitute for Problems in Mechanics, Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow, 119526, Russian Federation.
    Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, s. 2519-2535Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In this work we present results of development and application of Risk Oriented Accident Analysis framework (ROAAM+) to assessment of effectiveness of ex-vessel severe accident management strategy. In case of a core melt accident in Nordic type boiling water reactor (BWR) corium is released into a deep pool of water below reactor vessel to form a porous bed of debris. Energetic steam explosion or formation of non-coolable debris can threaten containment integrity. Both stochastic (aleatory) accident scenario and modeling (epistemic) uncertainties contribute to uncertainty. ROAAM+ framework is developed to simulate the whole accident progression The analysis starts from plant damage states determined in PSA Level-1 and continues with analysis of core degradation, vessel failure, melt release, steam explosion and debris bed formation and coolability. In order to achieve computational efficiency sufficient for extensive sensitivity, uncertainty, and risk analysis the surrogate modeling approach is used. In the paper we present results of the analysis aimed at quantification of uncertainty in the conditional containment failure probability. Specifically, we carry out sensitivity analysis using standalone and coupled models in order to identify the most influential scenario and modeling parameters for each sub-model. We assess the impact of the parameters on the prediction of the “load”, “capacity” and also failure probability. Then we quantify the effect of the most influential parameters on the failure probability. The results are presented using the failure domain approach and second order probability analysis, considering the uncertainty in distributions of the input parameters.

  • 20.
    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.

  • 21.
    Kööp, Kaspar
    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.
    Automated calibration and validationof RELAP5 input model against TALL-3D facility experimental dataInngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Validation of System Thermal Hydraulics (STH) codes against liquid metal facilities is necessary to increase confidence in designing and licensing of generation IV nuclear power systems. Manual input calibration and tuning against a single set of data can lead to bias in the result of the simulation towards specific system configuration and operation regime.In this work we demonstrate an approach to validation of the RELAP5 code, specifically, applicability of RELAP5 to model complex transients from forced to natural circulation in TALL-3D facility with Lead Bismuth Eutectic (LBE) coolant. We utilize an automated approach to (i) calibration of the input model using different experimental data and (ii) quantification of the modelling uncertainties. The automated approach is intended to reduce the effect of the user on the validation outcomes.Results from the calibrated model are compared against an experiment and uncertainty bounds presented. We discuss the results, provide recommendation to the modelling and provide conclusions on the applicability of the RELAP5 to simulation of different transients.

  • 22.
    Manickam, Louis
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    On the influence of water subcooling and melt jet parameters on debris formation2016Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 309, s. 265-276Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-jet facility by discharging binary-oxide mixtures of WO3-Bi2O3 and WO3-ZrO2 into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased sub cooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

  • 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. Papukchiev, A.
    et al.
    Geffray, C.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of the system thermal-hydraulics code ATHlet for the simulation of transient lead-bismuth eutectic flows2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, s. 810-822Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Within the European SESAME project, system thermal-hydraulics (STH), computational fluid dynamics (CFD) and coupled 1D-3D thermal-hydraulic simulations were carried out for Generation IV nuclear systems. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH participated in the project with activities related to the development and validation of CFD and coupled CFD-STH codes. The TALL-3D facility, operated by the KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant. A well-instrumented, partially heated test section with cylindrical form is installed in the primary circuit, which is domain of complex 3D flow and heat transfer phenomena. Three different experiments were calculated within a benchmark, organized by KTH: two with coupled programs and one with STH stand-alone code. This paper focuses on the analysis of the observed thermal-hydraulic flow phenomena during the TG03.S310.01 experiment and the comparison between ATHLET predictions and data.

  • 25. Papukchiev, A.
    et al.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Importance of conjugate heat transfer modeling in transient CFD simulations2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, s. 615-626Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Within the European SESAME project, system thermal-hydraulics (STH), computational fluid dynamics (CFD) and coupled 1D-3D thermal-hydraulic simulations are being carried out for Generation IV nuclear systems. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH participated in the project with activities related to the development and validation of CFD and coupled CFD-STH codes. The TALL-3D facility, operated by the KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant. A well-instrumented, partially heated test section with cylindrical form is installed in the primary circuit. It is the domain of complex 3D flow and heat transfer phenomena. The CFD analyses showed that only the consideration of the fluid domain is not sufficient for correct prediction of the thermal-hydraulic transient dynamics. Therefore, solid structures like test section walls, inner circular plate, test section heater and even the insulation were explicitly included in the CFD model. This allowed the consideration in the simulations of the heat conduction and the thermal inertia of these components. The paper focuses on the analysis of the observed thermal-hydraulic flow phenomena during the TG03.S301.04 experiment and the comparison between ANSYS CFX predictions and data. Moreover, the influence of conjugate heat transfer (CHT) on the numerical results is investigated and discussed.

  • 26.
    Phung, Viet-Anh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks2018Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, s. 461-476Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Characteristics of debris bed in the reactor vessel lower plenum affect corium interactions with the vessel structures, vessel failure and melt ejection mode. Melt release from the vessel determines conditions for ex-vessel accident progression and potential threats to containment integrity, i.e. steam explosion, debris bed formation and coolability. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles. Another problem is that code predictions can be quite sensitive to small variations of the input parameters and numerical discretization, due to the highly non-linear feedbacks and discrete thresholds employed in the code models.

     

    The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. Station black out scenario with a delayed power recovery is considered. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.

  • 27.
    Wang, Xicheng
    et al.
    KTH.
    Gallego-Marcos, Ignacio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Post-test calibration of the Effective Momentum Source (EMS) model for steam injection through multi-hole spargers.2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, s. 6176-6189Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Steam condensation in a large pool is often used in light water reactors to prevent containment overpressure. In boiling water reactors, steam from the primary system can be released into a pressure suppression pool (PSP) in normal operation and during accidents through multi-hole spargers to control the pressure in the reactor vessel. Steam injection into the pool can lead to the development of thermal stratification that affects (i) pressure suppression capacity of the pool, (ii) operation of the safety systems that use PSP as a source of water (e.g. emergency core cooling system and containment spray). Modeling of direct contact condensation of steam presents a challenge for contemporary codes. Therefore, Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. EMS defines the time-averaged effect of steam injection into the pool in terms of a momentum source. For multi-hole spargers, the momentum source requires models for (i) momentum induced by multi-holes steam injection, (ii) direction (vertical angle) of the induced momentum, and profile of velocity in (iii) vertical and (iv) azimuthal directions. Previous works on EMS model validation and sensitivity study against PPOOLEX and HYMERES PANDA pool tests suggest the importance of all these factors for accurate prediction of the pool mixing behaviour. All these parameters, except the velocity profile in the azimuthal direction, were measured in PANDA facility and in Separate Effect Facility (SEF) at Lappeenranta Institute of Technology. The goal of this work is to develop a model for the azimuthal profile of radial velocity (APV) of water induced by steam injection through multi-hole spargers in a pressure suppression pool of a Nordic Boiling Water Reactor (BWR). In previous work, it was assumed that the APV is the same as the radial velocity profile in vertical cross section (which was measured in PANDA experiments using PIV) and can be described by axisymmetric jet expansion model. In this paper, APV is defined as a separate model with own closure for the jet diffusion rate. The effect of the steam mass flow rate is taken into account in the APV and respective jet expansion factor according to the experimental observations. Finally, we compare the pool temperature evolution in the experiment and simulations with the EMS model.

  • 28.
    Wang, Xicheng
    et al.
    KTH.
    Gallego-Marcos, Ignacio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Pre-test analysis for HYMERES-2 PANDA tests series for steam injection into pool through spargers2019Inngår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, s. 6190-6203Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Steam condensation in a pool of water is often used in light water reactors. Steam injection provides sources of heat and momentum, which can lead to the development of thermal stratification or mixing of the pool. Modelling of the direct steam condensation is computationally challenging, especially considering the complex geometry of the pool and spargers and duration of transients. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed and implemented to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. These models are utilized in this work and are subject for further development. The goal of this work is to provide pre-test analysis to support the design and selection of the test conditions for the Pressure Suppression Pool (PSP) test series (OECD/HYMERES-2). This test series is aiming to provide data for the development and validation of the EMS model predictive capabilities for the PSP phenomena. Specifically, it was proposed to extend the experimental database with regards to the sparger design, pool depth and depth of sparger submergence. In order to maximize the value of obtained data for the model development and validation, the pre-test analysis of HYMERES-2 is carried out. Compared to the HYMERES-1 test series, the sparger elevation above pool bottom will be increased in order to study the effect of the distance between the sparger head and thermocline interface on the rate of erosion of the stratified layer. The Particle Image Velocimetry (PIV) will be used for the measurement of the azimuthal profile of flow radial velocity around the sparger. A large number of thermocouples are provided in the vertical direction to capture the transient location of the thermocline. Selection of specific values for the sparger elevation, location of the PIV window, number of thermocouples, and steam injection conditions are based on the analysis provided in this work. In the pre-test analysis we use the previous EHS/EMS models which show a good agreement with the HYMERES-1 test series.

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