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  • 1.
    Bandaru, S V Ravikumar
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Komlev, Andrei A.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sköld, Per
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Upward-facing multi-nozzle spray cooling experiments for external cooling of reactor pressure vessels2020Inngår i: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 163, artikkel-id 120516Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Cooling by water spray is a well-known technology that can reach significantly higher Critical Heat Flux (CHF) compared to other cooling methods. For the light water reactor safety, the in-vessel retention (IVR) by external reactor vessel cooling (ERVC) is a comprehensive severe accident management strategy to arrest and confine the corium in the lower head of the reactor pressure vessel. Heat fluxes up to 1.5 MW/m2 have already been assumed attainable in low-power nuclear reactors while cooling required in high-power reactors is expected to reach 2.5 MW/m2. Instead of reactor lower head flooding and relying on cooling due to natural convection, a viable and more efficient alternative is to spray the external surface of the vessel. Given all the advantages of spray cooling reported in the literature, a lab-scale experimental facility was built to validate the efficiency of multi-nozzle spray cooling of a downward-facing heated surface inclined at different angles up to 90o. The facility employed a 2×3 matrix of spray nozzles to cool the FeCrAl alloy foil with an effectively heated surface area of 96 cm2 using water as the coolant. Heat loads and surface inclinations were varied parameters in the test matrix. The results show that no significant variations in spray cooling performance concerning the inclination of the heated surface. A surface heat flux of 2.5 MW/m2 was achieved at every inclination of the downward-facing surface. The results also indicate that more uniform liquid film distribution could be obtained for some inclinations, which in turn leads to maintaining low surface temperature. The obtained surface heat flux margin by spray cooling indicates that it is feasible to adopt IVR-ERVC strategy for a large power reactor.

  • 2.
    Bandaru, S V Ravikumar
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Multi-nozzle spray cooling of a reactor pressure vessel steel plate for the application of ex-vessel cooling2021Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 375, artikkel-id 111101Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Spray cooling is a versatile technology for various cooling applications involving high surface heat fluxes. Experimental facility was built to study heat transfer performance of an upward multi-nozzle array of water sprays impacting a surface of heated plate made of reactor vessel grade steel. The effect of inclination angles of the steel surface on the cooling performance was investigated to assess heat transfer in complex semispherical/ semielliptical geometry of large reactor lower head and to address possible application of spray cooling in severe accident management (SAM) of light water reactors (LWRs) based on In-vessel melt retention with external reactor vessel cooling (IVR-ERVC). Joule heating of SA302B steel foil of 0.15 mm thickness and surface area of 96 cm2 enabled prototypic heat fluxes to be evacuated during reactor accident. A 2×3 array of full jet narrow-coned pressure-swirl spray nozzles was used to reproduce multi-nozzle cooling. The tests were conducted as a series of consequent steady states realized at stepwise increasing power and surface heat fluxes up to the maximum values of 29 kW and 2.97 MW/m2 limited in the specific facility design. Seven surface inclinations, between 0o and 90o were tested and no significant variations in spray cooling performance with the inclination of the heated surface was found. The results indicated a promising prospect of using a multi-nozzle array system for cooling of large surface area of reactor lower head. Much higher heat fluxes can be safely extracted by spray cooling in comparison with the critical heat fluxes that appeared at RPV water pool cooling at natural convection. The maximum value of heat flux at direct spray impact zones and its drop-off slightly from the center to the periphery of the spray cone was detected in the tests. The water flow rate and liquid subcooling significantly influenced maximum steel surface temperature but had no noticeable effects on surface temperature uniformity. The spray-to-spray interaction had no observable effects on local surface temperatures, however, the colliding zones where four spray cones have visible effects on local surface temperatures due to poor liquid momentum. The results also showed that more uniform liquid film distribution could be obtained for some inclinations because of improved liquid drainage, which in turn leads to maintaining low surface temperatures. 

  • 3.
    Bian, Boshen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Daniele, Dovizio
    Nuclear Research and Consultancy Group.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Direct numerical simulation of internally heated natural convection in a hemispherical geometryManuskript (preprint) (Annet (populærvitenskap, debatt, mm))
    Abstract [en]

    Internally heated (IH) natural convection can be found in nature, industrial processes, or during a severe accident in a light water reactor. In this accident scenario, the nuclear reactor core and some internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV) and interact with the remaining coolant. Subsequent re-heating and re-melting under decay and oxidation heat creates a transition from a debris bed to a molten pool. The molten pool, which can involve more than hundred tons of dangerously superheated oxidic and metallic liquids, imposes thermo-mechanical loads on the vessel wall that can lead to a thermal and/or structural failure of the vessel and subsequent release of radioactive materials to the reactor pit, and can possibly make its way to the environment. This study uses Direct Numerical Simulation (DNS) to investigate homogeneous IH molten pool convection in a hemispherical domain using Nek5000, an open-source spectral element code. With a Rayleigh number of 1.6×1011, the highest reached through DNS in this confined hemispherical geometry, and a Prandtl number of 0.5, which corresponds to a prototypic corium, the study provides detailed information on the thermo-fluid behaviour. The results show a turbulent flow with three distinct regions, consistent with the general flow observations from the BALI experiments. The study also presents detailed information on turbulence, such as turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance. Additionally, the study provides 3D heat flux distributions along the boundaries. The heat fluxes along the top boundary fluctuate due to the turbulent eddies in the vicinity, while along the curved boundary the heat fluxes increase nonlinearly from the bottom to the top.

    Fulltekst (pdf)
    fulltext
  • 4.
    Bian, Boshen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dovizio, Daniele
    bNuclear Research and Consultancy Group, the Kingdom of the Netherlands.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. Nuclear Futures Institute, Bangor University, United Kingdom.
    Direct numerical simulation of internally heated natural convection in a hemispherical geometry2024Inngår i: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 220, artikkel-id 124997Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Internally heated (IH) natural convection can be found in nature, industrial processes, or during a severe accident in a light water reactor. In this accident scenario, the nuclear reactor core and some internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV) and interact with the remaining coolant. Subsequent re-heating and re-melting under decay and oxidation heat creates a transition from a debris bed to a molten pool. The molten pool, which can involve more than hundred tons of dangerously superheated oxidic and metallic liquids, imposes thermo-mechanical loads on the vessel wall that can lead to a thermal and/or structural failure of the vessel and subsequent release of radioactive materials to the reactor pit, and can possibly make its way to the environment. This study uses Direct Numerical Simulation (DNS) to investigate homogeneous IH molten pool convection in a hemispherical domain using Nek5000, an open-source spectral element code. With a Rayleigh number of 1.6 × 1011, the highest reached through DNS in this confined hemispherical geometry, and a Prandtl number of 0.5, which corresponds to a prototypic corium, the study provides detailed information on the thermo-fluid behavior. The results show a turbulent flow with three distinct regions, consistent with the general flow observations from the BALI experiments. The study also presents detailed information on turbulence, such as turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance. Additionally, the study provides 3D heat flux distributions along the boundaries. The heat fluxes along the top boundary fluctuate due to the turbulent eddies in the vicinity, while along the curved boundary the heat fluxes increase nonlinearly from the bottom to the top.

  • 5.
    Bian, Boshen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gong, J.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Scalability of Nek5000 on High-Performance Computing Clusters Toward Direct Numerical Simulation of Molten Pool Convection2022Inngår i: Frontiers in Energy Research, E-ISSN 2296-598X, Vol. 10, artikkel-id 864821Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In a postulated severe accident, a molten pool with decay heat can form in the lower head of a reactor pressure vessel, threatening the vessel’s structural integrity. Natural convection in molten pools with extremely high Rayleigh (Ra) number is not yet fully understood as accurate simulation of the intense turbulence remains an outstanding challenge. Various models have been implemented in many studies, such as RANS (Reynolds-averaged Navier–Stokes), LES (large-eddy simulation), and DNS (direct numerical simulation). DNS can provide the most accurate results but at the expense of large computational resources. As the significant development of the HPC (high-performance computing) technology emerges, DNS becomes a more feasible method in molten pool simulations. Nek5000 is an open-source code for the simulation of incompressible flows, which is based on a high-order SEM (spectral element method) discretization strategy. Nek5000 has been performed on many supercomputing clusters, and the parallel performance of benchmarks can be useful for the estimation of computation budgets. In this work, we conducted scalability tests of Nek5000 on four different HPC clusters, namely, JUWELS (Atos Bullsquana X1000), Hawk (HPE Apollo 9000), ARCHER2 (HPE Cray EX), and Beskow (Cray XC40). The reference case is a DNS of molten pool convection in a hemispherical configuration with Ra = 1011, where the computational domain consisted of 391 million grid points. The objectives are (i) to determine if there is strong scalability of Nek5000 for the specific problem on the currently available systems and (ii) to explore the feasibility of obtaining DNS data for much higher Ra. We found super-linear speed-up up to 65536 MPI-rank on Hawk and ARCHER2 systems and around 8000 MPI-rank on JUWELS and Beskow systems. We achieved the best performance with the Hawk system with reasonably good results up to 131072 MPI-rank, which is attributed to the hypercube technique on its interconnection. Given the current HPC technology, it is feasible to obtain DNS data for Ra = 1012, but for cases higher than this, significant improvement in hardware and software HPC technology is necessary.

  • 6.
    Bian, Boshen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of DNS and RANS Approaches on Turbulent Natural Convection against the BALI-Metal ExperimentManuskript (preprint) (Annet vitenskapelig)
    Abstract [en]

    During severe accident scenarios in nuclear reactors, the core and internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV), where they interact with any remaining coolant. This process can lead to the formation of a stratified molten pool, which is also called corium. It consists of dangerously superheated oxidic and metallic liquids, which imposes thermo-mechanical loads on the vessel wall. Typically, the molten pool separates into distinct layers, with a lighter layer of metallic materials on top and a denser layer of oxides at the bottom. The metal layer acts as a heat sink, absorbing heat from the heat-generating oxide layer and conducting it towards the inner wall of the RPV. This concentrated heat load to the vessel is known as the focusing effect.

    This study conducts numerical simulations of the turbulent natural convection flow in a fluid layer undergoing both top and lateral cooling based on the BALI-Metal 8U experiment. Different methods were employed, including Direct Numerical Simulation (DNS) and three Reynolds-Averaged Navier-Stokes (RANS) models: k-ω SST, standard k-ε, and Reynolds stress equation model (RSM). The simulation results are compared with experimental data, and the RANS models are assessed using the DNS results. The results reveal that DNS is able to reproduce a two-distinct region flow structure similar to the experimental observations. The k-ω SST model shows similar flow patterns and turbulent kinetic energy (TKE) profile as the DNS results. Regarding the temperature field, all simulations overpredict temperature compared to the experimental data, with DNS providing the closest results. The turbulent heat flux (THF) result shows the RANS models are incapable of accurately modelling THF in turbulent natural convection flow. The heat flux analysis demonstrates that DNS achieved good agreement with experimental data in terms of heat flux distribution and energy balance, while the RANS models underestimate the focusing effect. Furthermore, DNS captures the transient maximum heat flux on the lateral cooling wall, which is higher than the time-averaged value, an important factor for estimating the focusing effect.

    Fulltekst (pdf)
    fulltext
  • 7.
    Bian, Boshen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dovizio, Daniele
    Nucl Res & Consultancy Grp NRG, Arnhem, Netherlands..
    Direct numerical simulation of molten pool convection in a 3D semicircular slice at different Prandtl numbers2022Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 393, s. 111772-, artikkel-id 111772Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this paper, a Direct Numerical Simulation (DNS) of an internally heated (IH) natural convection in a 3D semicircular slice molten pool is conducted using Nek5000, a CFD solver with spatial discretization based on the spectral element method. The mesh requirements in the bulk and boundary layers are both fulfilled using known correlations. A calculation of a simplified internally heated box is first established with an excellent agreement to existing data. Next, simulation of the 3D semi-circular is performed showing qualitative agreement with the general flow observations from the BALI experiments. The velocity field shows that the flow domain is divided into three regions, i.e., intensive turbulent eddies in the upper domain, weak flow motion in the lower domain, and the descending flow along the curved boundary. Correspondingly, the temperature field in the upper domain is relatively homogenous, while that in the lower domain is characterized by stratified layers. Further, the heat flux distribution along the boundaries shows that the heat fluxes fluctuate along the top wall due to turbulent eddies, and the heat fluxes at the curved wall increase nonlinearly from the bottom to the top. Finally, the influence of Prandtl number indicates that smaller Prandtl number will lead to more turbulence eddies, deeper descending flow, and more even redistribution of heat thereby lowering the maximum heat flux to the curved walls.

  • 8.
    Chen, Yangli
    et al.
    KTH.
    Zhang, Huimin
    KTH.
    Villanueva, Walter
    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.
    A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, s. 22-37Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

  • 9. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum2015Inngår i: International Conference on Nuclear Engineering, Proceedings, ICONE, JSME , 2015Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 10. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Extension of the MELCOR code for analysis of late in-vessel phase of a severe accident2015Inngår i: IYCE 2015 - Proceedings: 2015 5th International Youth Conference on Energy, IEEE conference proceedings, 2015Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 11.
    Do-Quang, Minh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Centra, Linné Flow Center, FLOW.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Centra, Linné Flow Center, FLOW.
    Singer-Loginova, Irina
    KTH, Skolan för teknikvetenskap (SCI), Centra, Linné Flow Center, FLOW.
    Amberg, Gustav
    KTH, Skolan för teknikvetenskap (SCI), Centra, Linné Flow Center, FLOW.
    Parallel adaptive computation of some time-dependent materials-related microstructural problems2007Inngår i: Bulletin of the Polish Academy of Sciences: Technical Sciences, ISSN 0239-7528, Vol. 55, nr 2, s. 229-237Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 12.
    Du, Yuxuan
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ravikumar Bandaru, Satya V.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Complementary Simulations to Determine Heat Transfer Coefficients and the Maximum Heat Flux in Multi-Nozzle Spray Cooling Experiments2022Inngår i: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME International , 2022, Vol. 5, artikkel-id V005T05A002Konferansepaper (Fagfellevurdert)
    Abstract [en]

    For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2×3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample's dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25 MWm2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material.

  • 13.
    Fichot, F.
    et al.
    IRSN, Cadarache, France. illanueva, W.; Bechta, S..
    Carenini, L.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    A revised methodology to assess in-vessel retention strategy for high-power reactors2018Inngår i: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7, The American Society of Mechanical Engineers , 2018, Vol. 7Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 14.
    Gallego Marcos, Ignacio
    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.
    Modelling of pool stratification and mixing induced by steam injectionthrough blowdown pipes2018Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 112, s. 624-639Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Containment overpressure is prevented in a Boiling Water Reactor (BWR) by condensing steam into thepressure suppression pool. Steam condensation is a source of heat and momentum. Competition betweenthese sources results in thermal stratification or mixing of the pool. The interplay between the sources isdetermined by the condensation regime, steam mass flow rate and pool dimensions. Thermal stratificationis a safety issue since it limits the condensing capacity of the pool and leads to higher containmentpressures in comparison to a completely mixed pool with the same average temperature. The EffectiveHeat Source (EHS) and Effective Momentum Source (EMS) models were previously developed for predictingthe macroscopic effect of steam injection and direct contact condensation phenomena on the developmentof stratification and mixing in the pool. The models provide the effective heat and momentumsources, depending on the condensation regimes. In this work we present further development of theEHS/EMS models and their implementation in the GOTHIC code for the analysis of steam injection intocontainment drywell and venting into the wetwell through the blowdown pipes. Based on thePPOOLEX experiments performed in Lappeenranta University of Technology (LUT), correlations arederived to estimate the steam condensation regime and effective heat and momentum sources as functionsof the pool and steam injection conditions. The focus is on the low steam mass flux regimes withcomplete condensation inside the blowdown pipe or chugging. Validation of the developed methodswas carried out against the PPOOLEX MIX-04 and MIX-06 tests, which showed a very good agreementbetween experimental and simulation data on the pool temperature distribution and containmentpressure.

  • 15.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kapulla, R.
    Paranjape, S.
    Paladino, D.
    Laine, J.
    Puustinen, M.
    Räsänen, A.
    Pyy, L.
    Kotro, E.
    Pool stratification and mixing during a steam injection through spargers: analysis of the PPOOLEX and PANDA experiments2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 337, s. 300-316Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 16.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kapulla, Ralf
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paranjape, Sidharth
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paladino, Domenico
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Laine, Jani
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Puustinen, Markku
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Rasanen, Antti
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Pyy, Lauri
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Kotro, Eetu
    Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 347, s. 67-85Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 17.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kapulla, Ralf
    Paul Scherrer Institute (PSI), Switzerland.
    Paranjape, Sidharth
    Paul Scherrer Institute (PSI), Switzerland.
    Paladino, Domenico
    Paul Scherrer Institute (PSI), Switzerland.
    Laine, Jani
    Lappeenranta University of Technology (LUT), Finland.
    Puustinen, Markku
    Lappeenranta University of Technology (LUT), Finland.
    Räsänen, Antti
    Lappeenranta University of Technology (LUT), Finland.
    Pyy, Lauri
    Lappeenranta University of Technology (LUT), Finland.
    Kotro, Eetu
    Lappeenranta University of Technology (LUT), Finland.
    Pool Stratification and Mixing Induced by Steam Injection through Spargers: CFD modelling of the PPOOLEX and PANDA experimentsManuskript (preprint) (Annet vitenskapelig)
    Abstract [en]

    Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models were implemented in the CFD code of ANSYS Fluent 17.0 and calibrated against the PPOOLEX and PANDA experiments with spargers analysed by the authors in [1] (Gallego-Marcos, I., et al., 2018). Modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the momentum magnitude showed that it varies between 0.2 to 1.2 times the steam momentum at the injection holes. Differences of this fraction between the PPOOLEX and PANDA simulations are discussed. Analysis of the calculated flow above the cold stratified layer shows that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.

  • 18.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Puustinen, Markku
    Lappeenranta University of Technology (LUT), Finland.
    Räsänen, Antti
    Lappeenranta University of Technology (LUT), Finland.
    Tielinen, Kimmo
    Lappeenranta University of Technology (LUT), Finland.
    Kotro, Eetu
    Lappeenranta University of Technology (LUT), Finland.
    Effective momentum induced by steam condensation in the oscillatory bubble regime2019Inngår i: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533, Vol. 350, s. 259-274Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The spargers used in Boiling Water Reactors (BWR) discharge steam from the primary coolant system into a pool of water. Direct steam condensation in subcooled water creates sources of heat and momentum determined by the condensation regimes, called “effective sources” in this work. Competition between the effective sources can result in thermally stratification or mixing of the pool. Thermal stratification is a safety concern in BWRs since it reduces the steam condensation and pressure suppression capacity of the pool. In this work, we present semi-empirical correlations to predict the effective momentum induced by steam condensation in the oscillatory bubble regime, relevant for the operation of spargers in BWRs. A Separate Effect Facility (SEF) was designed and built at LUT, Finland, in order to provide the necessary data. An empirical correlation for the effective momentum as a function of the Jakob number is proposed. The Kelvin Impulse theory was also applied to estimate the effective momentum based on information about the bubble dynamics. To do this, new correlations for the bubble collapse frequencies, maximum bubble radius, velocities, pressure gradient and heat transfer coefficient are proposed and compared to available data from the literature. The effective momentum induced by sonic steam jets appears to be constant in a wide range of studied Jakob number. However, further experimental data is necessary at larger Jakob numbers and steam mass fluxes.

  • 19.
    Gallego-Marcos, Ignacio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kapulla, R
    Paranjape, S
    Paladino, D
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Modeling of Thermal Stratification and Mixing Induced by Steam Injection Through Spargers Into a Large Water Pool2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The pressure suppression pool of a Boiling Water Reactor (BWR) is designed to protect the containment from over pressure by condensing steam. Under certain steam injection conditions, thermal stratification can develop in the pool and significantly reduce its pressure suppression capacity. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers, which are multi-hole injection pipes connecting the main steam lines to the wetwell pool. The aim of the model is to predict the global pool behavior. Effective Heat and Momentum Sources (EHS/EMS) approach is used to model time averaged effects of small scale direct contact condensation phenomena on the large scale pool circulation. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments was done to address the most important physical phenomena that can occur in plant scale. The results show that the global pool behavior can be predicted using the Standard Gradient Diffusion Hypothesis (SGDH) in k-Omega turbulence model.

  • 20.
    Gallego-Marcos, Ignacio
    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.
    Modeling of Thermal Stratification and Mixing in a Pressure Suppression Pool Using GOTHIC2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it can lead to higher containment pressures than in completely mixed conditions. The thermal hydraulic code of GOTHIC offers a very suitable platform to simulate the pool and containment behavior during a long term accident. However, for a computationally efficient code such as GOTHIC, direct contact condensation cannot be resolved accurately enough to obtain a good estimation of the momentum induced by the condensing steam, and thus, to predict the pool behaviour. In this paper, we present how to implement the previously validated Effective Heat Source (EHS) and Effective Momentum Source (EMS) models, developed for pool analysis during a steam injection, in GOTHIC. The implementation was done using control variables and Dynamically Linked Libraries (DLL). A time averaging model to minimize the effect of the numerical oscillations appearing in GOTHIC when steam is injected into the pool is also proposed.

  • 21.
    Gallego-Marcos, Ignacio
    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.
    Possibility of Air Ingress into a BWR Containment during a LOCA in case of Activation of Containment Venting System2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The pressure relief systems installed in BWRs protect the containment from overpressure in case of a Loss of Coolant Accident (LOCA). This paper analyzes the possibility of air ingress, which can cause hydrogen burn, through the rupture disks of the filtered and non-filtered venting systems. Two scenarios were considered: a LOCA without SBO (Station Blackout) and a LOCA with SBO. The thermal-hydraulic code GOTHIC® was used with 3D models of the drywell and wetwell of a Nordic-type BWR. In the LOCA event, we found no activation of the rupture disks within the considered transient simulation. Moreover, the containment spray ensured a low pressure in the drywell and induced a continuous mixing of the wetwell pool. In the LOCA with SBO event, the development of thermal stratification in the wetwell pool accelerated the pressure increase in the drywell, which led to activation of the rupture disk of the filtered venting system. However, no air ingress through the vent was found during the depressurization of the containment, and hence no risk of hydrogen burn under the given assumptions.

  • 22.
    Gallego-Marcos, Ignacio
    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.
    Scaling and CFD Modelling of the Pool Experiments with Spargers Performed in the PANDA Facility2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it reduces its cooling capability and leads to higher containment pressures than in completely mixed conditions. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers. The model provides the time averaged heat and momentum transferred from the steam condensation to the large scale pool circulation. Small scale phenomena such as direct contact condensation is not resolved, only its effect on the pool behaviour. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments, done to preserve the most important physical phenomena occurring in plant scale is also presented in the paper. The results show that the model is able to predict well the global pool behavior. However, flow instabilities were observed to induce a sudden mixing of the upper part of the stratified layer during the transition from the stratification to the mixing phases. This led to a faster erosion of the layer than in the experiment. Simulations done with 2D and 3D meshes and scale adaptive turbulence models were performed to clarify this issue and are presented in the paper.

  • 23.
    Gallego-Marcos, Ignacio
    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.
    Water Distribution in a Nordic BWR Containment During a LOCA2016Inngår i: 2016 International Congress on Advances in Nuclear Power Plants, ICAPP 2016, 2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    During a main steam line break in a Boiling Water Reactor (BWR) the pressure suppression pool is used as a water source for the Emergency Core Cooling System (ECCS) and the Containment Spray (CS). These systems drain water from the pool through strainers, which are long perforated plates or cylinders submerged to a certain depth. Proper functioning of the ECCS and the CS must be ensured to maintain the water inventory in the vessel and to limit the containment pressure. However, if the liquid level in the suppression pool goes below the level of the strainers intake, the operators would be forced to stop their pumps. The liquid level in the suppression pool can be reduced when a significant fraction of ECCS and CS flow is relocated to the lower drywell. In this work, we use the thermal-hydraulic code GOTHIC to simulate the containment evolution during a main steam line break inside the biological shield. The containment volumes and their connections were modeled with 2D and 3D volumes. With this model, scenarios considering different operational conditions were assessed: (i) full capacity of all the safety systems, (ii) half capacity of all the safety systems, (iii) ECCS stops injecting water after a certain liquid level is restored in the vessel, and (iv) the pipes used to drain water from the suppression pool and flood the lower drywell are partially or totally clogged in different directions. The results showed that there is a risk of an early shut down of the ECCS and CS systems in the case of main steam line break inside the biological shield. It was observed that when the ECCS provided a continuous water injection into the vessel, the water spilled through the break into the biological shield flowed downwards driven by gravity and went directly into the lower drywell. This caused a fast decrease in the liquid level of the suppression pool, which led to an uncovery of the ECCS and CS strainers about 2000 s after the break. The activation at 1800 s of the flooding of the lower drywell led to a backward flow, from the lower drywell to the suppression pool, since at that time the liquid level in the suppression pool was lower than in the lower drywell. However, this backward flow was not enough to maintain the liquid level in the suppression pool, which continued to decrease. In the case where the pipes used for the flooding were clogged in the direction of the suppression pool, uncovery of the strainers was observed even earlier.

  • 24.
    Galusin, Sergey
    et al.
    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.
    Development of Core Relocation Surrogate Model for Prediction of Debris Properties in Lower Plenum of a Nordic BWR2016Inngår i: NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Gyeongju, Korea, October 9-13, 2016. Paper N11P1234, NUTHOS-11 , 2016, artikkel-id N11P1234Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. A consistent treatment of these uncertainties requires Integrated Deterministic Probabilistic Safety Analysis (IDPSA). We employ the concepts and approaches described in Risk Oriented Accident Analysis Methodology (ROAAM) for development of a probabilistic framework (ROAAM+) that is based on extensive uncertainty and sensitivity analysis in risk quantification. Direct application of such fine-resolution models for extensive sensitivity and uncertainty analysis is often unaffordable. We use “surrogate models” (SMs) that provide computationally efficient approximations for the FMs. In this work we demonstrate an approach to the development of Core relocation SM based on the MELCOR code as the full model (FM). We discuss the development of the database of the FM solutions, data mining and post-processing of the results for SM development. Extensive sensitivity and uncertainty analysis is carried out using the FM and implications of the analysis are discussed in detail. We demonstrate how the connection between different stages of severe accident progression is made in ROAAM+ framework for Nordic BWRs.

    Fulltekst (pdf)
    fulltext
  • 25.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi-Thanh
    Effect of Corium Non-homogeneity on Nordic BWR Vessel Failure Mode and Timing2015Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Corium melt fragmentation and cooling in a deep pool of water under reactor pressure vessel are employed as severe accident mitigation strategy in a Nordic-type BWR. Core debris relocated to the lower head inflict significant thermal and mechanical loads on the vessel structures. The mode and timing of the vessel failure, mass and superheat of the ejected melt determine ex-vessel accident progression and risks of steam explosion and formation of a non-coolable debris bed. In this work we consider the effect of in-vessel debris non-homogeneity on the mode of vessel failure. The heat-up, re-melting, melt pool formation, and heat transfer of the debris bed are predicted with the Phase-change Effective Convectivity Model (PECM) implemented in FLUENT® code. Then the obtained thermal load on the vessel wall and structures is used as boundary conditions for a thermo-structural analysis of the BWR lower head using the ANSYS® code. In this paper, a corium debris bed is considered inside vessel lower head inducing thermal load on the wall and structures. The debris bed thermal properties axial distribution is taken as a function of material composition, which is extracted from MELCOR® simulations of core failure and debris bed formation inside the lower plenum. A flat and a concave configuration of the debris bed are considered and results of simulations are compared with those for a homogenous debris bed of the same mass-averaged thermal properties.

  • 26.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The Effect of Internal Pressure and Debris Bed Thermal Properties on BWR Vessel Lower Head Failure and Timing2013Inngår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 27.
    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.

  • 28.
    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.

  • 29.
    Guo, Qiang
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Manickam, Louis
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Yu, Peng
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    A design study on an aerodynamic levitation system for droplet preparation in steam explosion experiment2019Inngår i: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In order to investigate the mechanisms of steam explosion which may occur during a severe accident of light water reactors (LWRs), the MISTEE facility was developed at Royal Institute Technology (KTH) to visualize the micro interactions of steam explosion when a single molten droplet was falling into a water pool. For preparation of a molten droplet, an aerodynamic levitation system was proposed to prevent the droplet from falling out of the crucible during heating in an induction furnace by injecting argon gas through a purging line connected to the bottom nozzle of the crucible. To support the design of such levitation system, a numerical simulation of the aerodynamic levitation system was performed using the CFD code ANASYS FLUENT v16.2. The problem was simplified as adiabatic two-phase flow dynamics in a 2-D axisymmetric geometry. The VOF method is employed to track the interface of two phases (liquid metal and argon gas), and the SST k-omega model was chosen for turbulence. Various characteristics of droplet dynamics in incorporated with argon gas flowrates through the crucible were examined in the numerical simulation. The simulation results suggested there exists an optimal range of argon gas flowrate for levitating a coherent shape of droplet in the crucible. The wall adhesion had a considerable effect on initiating the levitation of the droplet, which means the properties of the inside surface of the crucible may play an important role in the levitation and discharge of the droplet. Proof-of-concept tests were carried out on the prototype of the design, and it was confirmed that the levitation system was able to fulfill its function, i.e., to keep the droplet in the crucible during its melting process, although the actual argon gas flowrates for levitation was higher than the predicted ones, probably due to the leakage of flow path and heat transfer which were not considered in the simulation. Generally speaking, the numerical simulation did not only help understand the hydrodynamic characteristics of the levitation system, but also provided insights on operational control and improvement of the system.

  • 30.
    Hoseyni, Seyed Mohsen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Komlev, Andrei A.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sköld, Per
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Akiba, Miyuki
    Secretariat Nucl Regulat Author S NRA R, Regulatory Stand & Res Dept, Tokyo, Japan..
    Hotta, Akitoshi
    Secretariat Nucl Regulat Author S NRA R, Regulatory Stand & Res Dept, Tokyo, Japan..
    Metallic melt infiltration in preheated debris bed and the effect of solidification2021Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 379, s. 111229-, artikkel-id 111229Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The re-melting of multi-component debris is important for both in-vessel and ex-vessel phases of severe accident progression in nuclear power plants. However, current knowledge is limited with respect to understanding the associated complex phenomena and their interactions. In this paper, the phenomenon of melt infiltration through a porous debris bed with and without solidification is examined by synthesizing the data obtained from ongoing experimental research (REMCOD facility). In this regard, results obtained from 12 experiments are analyzed. Eight tests were conducted for melt infiltration through debris at temperatures above solidification. At this condition, two flow regimes are identified for the melt flow inside the hot porous debris, which is initially dominated by capillary forces and hydrostatic head and then later by the gravity forces. In addition, 4 tests were performed for melt penetration into cold debris where melt infiltration is limited by solidification. It was found that the depth of penetration is correlated with the difference between "sensible heat of melt" and "the amount of heat required to heat the bed up to the melting point of specific melt composition."

  • 31.
    Hoseyni, Seyed Mohsen
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Melt infiltration through porous debris at temperatures above Solidification: Validation of analytical model2021Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 161, s. 108435-, artikkel-id 108435Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper investigates the dynamics of melt infiltration through a preheated porous debris bed which is of importance to severe accident modeling in nuclear power plants. Proper understanding of the flow physics and affecting parameters is needed to define flow regime(s) according to combination of the driving forces, i.e. capillary and gravity. A model development and validation therefore should consider various effects and competing mechanisms. After a careful study of the governing equations and scaling rules, a known analytical model is validated against existing experimental data from REMCOD experiment. The predictions of this model are in good agreement with the experimental data. Furthermore, a global sensitivity analysis identifies the most influential parameters and reveals the need for further experiments with different range of affecting parameters. The results underline the importance of permeability as the most influential parameter.

  • 32.
    Hossny, Karim
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. Nuclear Futures Institute, School of Computer Science and Electronic Engineering, Bangor University, Bangor, LL57 1UT, UK.
    Wang, Hongdi
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Distinctive physical insights driven from machine learning modelling of nuclear power plant severe accident scenario propagation2023Inngår i: Scientific Reports, E-ISSN 2045-2322, Vol. 13, nr 1, artikkel-id 930Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The severe accident scenario propagation studies of nuclear power plants (NPPs) have been one of the most critical factors in deploying nuclear power for decades. During an NPP accident, the accident scenario can change during its propagation from the initiating event to a series of accident sub-scenarios. Hence, having time-wise updated information about the current type of accident sub-scenario can help plant operators mitigate the accident propagation and underlying consequences. In this work, we demonstrate the capability of machine learning (Decision Tree) to help researchers and design engineers in finding distinctive physical insights between four different types of accident scenarios based on the pressure vessel's maximum external surface temperature at a particular time. Although the four accidents we included in this study are considered some of the most extensively studied NPPs accident scenarios for decades, our findings shows that decision tree classification could define remarkable distinct differences between them with reliable statistical confidence.

  • 33. Hotta, A.
    et al.
    Akiba, M.
    Morita, A.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Tillämpad fysik, Nanostrukturfysik. KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Centra, Linné Flow Center, FLOW. KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Komlev, Andrei A.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Hoseyni, Seyed Mohsen
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sköld, Per
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Matsumoto, T.
    Sugiyama, T.
    Buck, M.
    Experimental and Analytical Investigation of Formation and Cooling Phenomena in High Temperature Debris Bed2019Inngår i: Journal of Nuclear Science and Technology, ISSN 0022-3131, E-ISSN 1881-1248Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 34.
    Hua, Li
    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.
    Validation of Effective Momentum and Heat Flux Models for Stratification and Mixing in a Water Pool2013Rapport (Annet vitenskapelig)
    Abstract [en]

    The pressure suppression pool is the most important feature of the pressure suppression system in a Boiling Water Reactor (BWR) that acts primarily as a passive heat sink during a loss of coolant accident (LOCA) or when the reactor is isolated from the main heat sink. The steam injection into the pool through the blowdown pipes can lead to short term dynamic phenomena and long term thermal transient in the pool. The development of thermal stratification or mixing in the pool is a transient phenomenon that can influence the pool's pressure suppression capacity. Different condensation regimes depending on the pool's bulk temperature and steam flow rates determine the onset of thermal stratification or erosion of stratified layers. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (i) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (ii) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale modeling, and (c) validate our proposed models. The data from PPOOLEX STR-06, STR-09 and STR-10 tests are used for validation of the EHS and EMS models in this work. We found that estimations of the amplitude and frequency based on available experimental data from PPOOLEX experiments STR-06, STR-09, and STR-10 have too large uncertainties due to poor space and time resolution of the temperature measurements in the blowdown pipe. Nevertheless, the results demonstrated that simulations with variable effective momentum which is selected within the experimental uncertainty have provided reasonable agreement with test data on transient temperature distribution in the pool. In order to reduce uncertainty in both experimental data and EHS/EMS modeling, additional tests and modifications to the experimental procedures and measurements system in the PPOOLEX facility were proposed. Pre-test simulations were performed to aid in determining experimental conditions and procedures. Then, a new series of PPOOLEX experimental tests were carried out. A validation of EHS/EMS models against MIX-01 test is presented in this report. The results show that the clearing phase predicted with 3D drywell can match the experiment very well. The thermal stratification and mixing in MIX-01 is also well predicted in the simulation.

  • 35.
    Hultgren, Ante
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gallego-Marcos, Ignacio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Simulation of Large Scale Erosion of a Stratified Helium Layer by a Vertical Air Jet using the GOTHIC Code2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In case of a severe core degradation in a Light Water Reactor (LWR), significant amount of hydrogen can be produced posing a risk of hydrogen burning and detonation. Reliable prediction of hydrogen build-up, stratification, and mixing in the containment is of paramount importance since the phenomena affect hydrogen distribution in the containment. In this paper, we present a modeling approach using the GOTHIC code. The simulation results were compared against experimental data from the ST1-7 experiment performed in the PANDA facility at the Paul Scherrer Institute (PSI). The ST1-7 experiment consists of an air jet impingement onto a stratified helium layer. The modelling approach uses coupled volumes to introduce in each region of the computational domain (i) adequate mesh resolutions to resolve the gradients of the flow and (ii) appropriate turbulence models in order to resolve locally dominant flow structures. With the adaptive mesh, only about 7400 cells for the 2 PANDA vessels (4 m diameter by 8 m in height cylinders with an interconnecting pipe) is enough to provide reasonably accurate results. We found that using the k-epsilon standard model for the jet region and the mixing length model for the rest of the domain, has provided remarkably good agreement with the experimental data. The erosion of the helium stratified layer before and after the air injection is discussed in detail.

  • 36. Jeltsov, Marti
    et al.
    Cadinu, F.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    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.
    An Approach to Validation of Coupled CFD and System Thermal-Hydraulics Codes.2011Konferansepaper (Fagfellevurdert)
  • 37.
    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.
    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.
    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)
  • 38.
    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.
    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.
    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)
  • 39.
    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)
  • 40.
    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 domain overlapping STH/CFD coupling approach for analysis of heavy liquid metal thermal hydraulics in TALL-3D experiment2012Konferansepaper (Fagfellevurdert)
  • 41.
    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.

    Fulltekst (pdf)
    fulltext
  • 42.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    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.

  • 43.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    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.

  • 44.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Seismic sloshing effects in lead-cooled fast reactors2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, s. 99-110Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 45.
    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.

  • 46.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Yakush, Sergey
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. AlbaNova University Center.
    Phung, Viet-Anh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Nam
    A framework for assessment of severe accident management effectiveness in Nordic BWR plants2014Inngår i: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

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

  • 47.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Agglomeration and size distribution of debris in DEFOR-A experiments with Bi2O3-WO3 corium simulant melt2013Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 263, s. 284-295Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

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

  • 48.
    Kudinov, Pavel
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The DEFOR-A Experiment on Fraction of Agglomerated Debris as a Function of Water Pool Depth2010Inngår i: The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), 2010Konferansepaper (Fagfellevurdert)
  • 49.
    Kudinov, Pavel
    et al.
    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.
    Fraction of Agglomerated Debris in Experiment on Pouring of High Temperature Melt in Water2010Inngår i: Proc. 8th International TopicalMeeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), 2010Konferansepaper (Fagfellevurdert)
  • 50.
    Li, Hua
    et al.
    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.
    Condensation, Stratification and Mixing in aBWR Supression Pool2010Rapport (Annet vitenskapelig)
12 1 - 50 of 93
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