| 1. |
- Galushin, Sergey, et al.
(författare)
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Treatment of Phenomenological Uncertainties in Level 2 PSA for Nordic BWR Using Risk Oriented Accident Analysis Methodology
- 2022
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Ingår i: Probabilistic Safety Assessment and Management, PSAM 2022. - : International Association for Probablistic Safety Assessment and Management (IAPSAM).
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Konferensbidrag (refereegranskat)abstract
- A comprehensive and robust assessment of phenomenological uncertainties is a challenge for the current real-life PSA L2 applications, since such uncertainty is majorly driven by physical phenomena and timing of events. Typically, the static PSA models are built on a pre-determined set of scenario parameters to describe the accident progression sequence and use a limited number of simulations in the underlying deterministic analysis to evaluate the consequences. The Risk Oriented Accident Analysis Methodology (ROAAM+) has been developed to enable consistent and comprehensive treatment of both epistemic and aleatory sources of uncertainty in risk quantification. The framework is comprised of a set of deterministic models that simulate different stages of the accident progression, and a probabilistic platform that performs quantification of the uncertainty in conditional containment failure probability. This information is used for enhanced modeling in the PSA-L2 for improved definition of sequences, where information from the ROAAM is used to refine PSA model resolution regarding risk important accident scenario parameters, that can be modelled within the PSA. This work presents an example of application of the dynamic approach in a large-scale PSA model and demonstrate the integration of the ROAAM+ results in the PSA model.
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| 2. |
- Wong, Kin Wing, et al.
(författare)
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CFD STUDIES OF SEPARATE EFFECT FLOW ACCELERATED CORROSION AND EROSION (SEFACE) FACILITY FOR HEAVY LIQUID METAL
- 2023
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Ingår i: Proceedings of the 30th International Conference on Nuclear Engineering "Nuclear, Thermal, and Renewables. - : American Society of Mechanical Engineers (ASME).
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Konferensbidrag (refereegranskat)abstract
- Long-term material compatibility in heavy liquid metal (HLM) remains a challenge for the successful deployment of HLM-based technologies. Flow-accelerated corrosion and erosion (FACE) phenomena can lead to continual material deterioration, which needs to be considered throughout the reactor design stage. Nonetheless, known experimental data are inadequate to cover all the prototypical flow regimes during LFR's operation. Modelling of the FAC/FACE phenomena remains mostly in lumped parameter/subchannel scales, where the FAC model is coupled to the bulk flow of the pipe or subchannel. These methodologies might produce a sufficient prediction for the core internals; however, this might not be suitable for the pump impeller due to comparatively greater relative velocity and the occurrence of transient flow patterns near the rotating impeller. To establish an understanding of the connection between turbulence and FACE, the liquid lead-based Separate Effect Flow Accelerated Corrosion and Erosion (SEFACE) facility is currently under design at KTH in the framework of the Sustainable Nuclear Energy Research In Sweden (SUNRISE) project. SEFACE attempts to investigate FACE phenomena in the liquid lead and produce quantifiable validation data for model development. The paper divides itself into two parts. Part I refers to the study of operational conditions in SEFACE via Reynolds Averaged Navier Stokes (RANS) simulation, while Part II deals with the recent attempt on modelling time-dependent flow shear on rotating disks based on large eddy simulation (LES). The paper begins with a brief review of prior studies on flow-accelerated corrosion. Following that, the SEFACE facility's design concept is laid out considering several physical and operational constraints. A periodic wedge of the SEFACE test chamber is chosen to examine the facility's time-averaged behaviour. The k-ω shear stress transport (SST) model was employed for the simulations. The torque prediction on the rotating disk system is verified with the empirical frictional factor prediction. The latest hydrodynamic design enables SEFACE to be spun at 1200 revolutions per minute (corresponding to a maximum velocity of 21 m/s) without causing free surface deformation or excessive pressure. SEFACE permits the collecting of experimental data under the effect of various relative velocities in a single experiment round. The second part of the paper focuses on a recent attempt to determine the wall shear stress distribution on a rotating disk using wall-modelled large eddy simulation (WMLES S-Omega). The obtained amplitude and frequency of wall shear stress fluctuations will aid model development in future.
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| 3. |
- Isaev, S.A., et al.
(författare)
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Numerical analysis of the jet-vortex pattern of flow in a rectangular trench
- 2003
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Ingår i: Journal of Engineering Physics and Thermophysics. - 1062-0125 .- 1573-871X. ; 76:2, s. 257-265
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Tidskriftsartikel (refereegranskat)abstract
- Numerical simulation of the vortex structure of three-dimensional laminar flow in a rectangular trench of square cross section has been carried out on the basis of the finite-volume solution of steady-state Navier-Stokes equations.
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| 4. |
- Konovalenko, Alexander, et al.
(författare)
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Experiments and modeling of particulate debris spreading in a pool
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 8055-8068
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Konferensbidrag (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase turbulent flows in the pool serve as a source of mechanical energy which can affect the initial geometry as well as dynamically change the shape of already formed debris bed. The main goal of this work is to provide experimental data on spreading of solid particles in the pool by large scale two-phase flow structures induced by gas injection from the bottom. These data are necessary for development and validation of predictive capabilities of computer codes allowing numerical modeling of the debris bed formation at prototypic severe accident conditions. In PDS-P experiments air injection at the bottom of the test section is employed in order to create large scale flow in the pool. The test section is constructed as a rectangular tank. It has close to 2D geometry with fixed width (72 mm), variable length (up to 1.6 m) and allows water filling depth of up to 1 m. The variable pool length and depth allows formation of the different in size and pattern two-phase circulating flows. Experimental conditions such as gas-phase flow rate and particle properties (density and size) are scaled to maintain relevancy to the prototypic accident conditions. The average void fraction in the pool is determined by video recording and image processing. Particles are supplied from the top of the facility above the water surface. In the separate-effect studies of the influence of two-phase currents on particle trajectories and bed formation, a low particle flow rate is required in order to minimize or completely exclude particle-particle interaction. Results of several series of PDS-P (Particulate Debris Spreading in the Pool) reported in this paper are analyzed analytically. The preliminary scaling approach is proposed and has good agreement with experimental findings.
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| 5. |
- Kudinov, Pavel, et al.
(författare)
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Experimental investigation of debris bed agglomeration and particle size distribution using W03-ZR02 melt
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 8046-8054
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Konferensbidrag (refereegranskat)abstract
- Nordic BWR severe accident management strategy employs reactor cavity flooding to terminate ex-vessel accident progression. Corium melt released from the reactor pressure vessel is expected to fragment and form a porous debris bed. Success of the SAM strategy is contingent upon possibility to remove the decay heat generated in the debris bed by natural circulation of the coolant. Properties of the debris bed such as particle size, porosity and shape of the bed determine resistance for the coolant flow and thus dryout heat flux. Agglomeration of incompletely solidified debris can create additional obstacles for coolant circulation and thus reduce debris coolability margin. The goal of DEFOR (debris bed formation) experimental work is to provide data necessary for the development of analytical models and approaches for prediction of debris bed formation and agglomeration phenomena. Different corium simulant materials are used in the experiments. Liquid melt jet fragmentation and debris bed formation are considered at different conditions such as melt release (jet diameter, free fall height, etc.), melt superheat, water subcooling and water pool depth. A series of confirmatory DEFOR-A experiments has been carried out with ZrO2-WO3 simulant material. The data on particle size distribution, debris bed porosity and agglomeration is in good agreement with the previous DEFOR-S, DEFOR-A and FARO tests. On average, larger particles were obtained with ZrO2-WO3 melt than with previously used Bi2O3-WO3, size distributions for both melt simulant materials are within the ranges of size distributions observed in FARO tests. The difference between particle sizes in the tests with free falling jets was found to be insignificant. There is a tendency to form slightly larger particles only in the tests with submerged nozzles where melt is released under water with initially small jet velocity. Initial jet velocity also seems to have no visible effect on the fraction of agglomerated debris.
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| 6. |
- Kumar, Ranjan, et al.
(författare)
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Dynamic Hybrid Reliability Studies of a Decay Heat Removal System
- 2015
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Ingår i: International Journal of Reliability, Quality and Safety Engineering (IJRQSE). - : World Scientific. - 0218-5393. ; 22:4
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Tidskriftsartikel (refereegranskat)abstract
- Some critical safety systems exhibit the characteristics of hybrid stochastic class whose performance depends on the dynamic interactions of deterministic variables of physical phenomena and probabilistic variables of system failures. However, conventional probabilistic safety assessment (PSA) method involves static event and linked fault tree analysis and does not capture the dynamic interactions of such hybrid stochastic systems. Additionally, the existing dynamic PSA methods do not consider any repair possibility of some failed components during safety assessment. To address these issues, this paper presents a dynamic hybrid reliability assessment scheme for performance studies of repairable nuclear safety systems during a mission time. This scheme combines the features of reliability block diagram (RBD) for system compositions and partial differential equations for system physics using a customized stochastic hybrid automata tool implemented on Python platform. A case study of decay heat removal (DHR) systems has been performed using the introduced scheme. The impacts of failure rates and repair rates on sodium temperature evolution over a mission time have been analyzed. The results provide useful safety insights in mission safety tests of DHR systems. In sum, this work advances the dynamic safety assessment approach for complex system designs including nuclear power plants.
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| 7. |
- Mickus, Ignas, et al.
(författare)
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Development of tall-3d test matrix for APROS code validation
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 4562-4575
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Konferensbidrag (refereegranskat)abstract
- APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.
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| 8. |
- Moreau, V., et al.
(författare)
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Pool CFD modelling : lessons from the SESAME project
- 2019
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Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 355
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Tidskriftsartikel (refereegranskat)abstract
- The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.
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| 9. |
- Papukchiev, Angel, et al.
(författare)
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On the need for conjugate heat transfer modeling in transient CFD simulations
- 2020
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 367
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Tidskriftsartikel (refereegranskat)abstract
- Within the European SESAME project, system thermal-hydraulics (STH), computational fluid dynamics (CFD) and coupled 1D-3D thermal-hydraulic simulations were carried out for Generation IV nuclear reactor systems. The Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) gGmbH participated in the project with activities related to the development and validation of coupled STH-CFD codes. The TALL-3D facility, operated by the Royal Institute of Technology (KTH) in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant. A well-instrumented, partially heated test section with cylindrical form is installed in the primary circuit. It is the domain of complex 3D flow and heat transfer phenomena. The CFD analyses showed that only the consideration of the fluid domain is not sufficient for the correct prediction of the thermal-hydraulic transient dynamics. Therefore, solid structures like the test section walls, inner circular plate, test section heater, and the insulation were explicitly included in the CFD model. This allowed the consideration in the simulations of the heat conduction and the thermal inertia of these components. The paper focuses on the analysis of the observed thermal-hydraulic flow phenomena in two experimental cases, and the comparison between CFD predictions and data. Both deal with the transition from forced to natural circulation. In the first experiment the test section heater is switched on, while in the second the heat is generated only in the main heater component. The influence of conjugate heat transfer (CHT) on the numerical results is investigated and discussed in detail.
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| 10. |
- Tran, C. -T, et al.
(författare)
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A study on transient heat transfer of the EU-ABWR external core catcher using the Phase-change Effective Convectivity Model
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - : American Nuclear Society. - 9781510811843 ; , s. 6821-6834
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Konferensbidrag (refereegranskat)abstract
- In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation HI and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and procedures are evaluated based on results of PECM simulation and severe accident analysis.
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