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  • 1.
    Anglart, Henryk
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Bergagio, Mattia
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Experimental and numerical investigations of wall temperature fluctuations due to thermal mixing in an annulus2016Conference paper (Refereed)
    Abstract [en]

    Wall temperature fluctuations during thermal mixing of water in an annular test section have been measured and numerically predicted. The characteristics of the temperature fluctuations, such as their amplitudes and frequencies, are closely related to a premature structural failure due to the thermal fatigue. The goal of the present work has been to obtain experimental data on the convective heat transfer in presence of thermal mixing and use the data for validation of computational codes. During the experiments, two water streams at significantly different temperatures and at pressure 7.2 MPa are mixing in an annular test section, causing significant fluctuations of temperatures in walls surrounding the mixing zone. In parallel to experiments, the analyses of water mixing and of the resulting wall temperature fluctuations have been carried out using the Large Eddy Simulations (LES) with conjugate heat transfer approach. A similar behavior of temperature fluctuations has been observed in experiments and calculations. In particular, it has been both calculated and measured that the wall temperature spectrum varies at different locations in the test section and the dominant frequencies of fluctuations for the case presented in the paper are in the range of 0.1 to 0.2 Hz.

  • 2.
    Bergagio, Mattia
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Fan, Wenyuan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, Superseded Departments (pre-2005), Physics. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Large eddy simulation of thermal mixing with conjugate heat transfer at BWR operating conditionsManuscript (preprint) (Other academic)
    Abstract [en]

    Thermal fatigue occurs in most metals under cyclic heat loads and can threaten the structural integrity of metal parts. Detailed knowledge of these loads is of utter importance to prevent such issues. In this study, a large eddy simulation (LES) with wall-adapting local eddy viscosity (WALE) subgrid model is performed to better understand turbulent thermal mixing in an annulus with a pair of opposing cold inlets at a low axial level (z = 0.15 m) and with a pair of opposing hot inlets at a higher axial level (z = 0.80 m). Each inlet pair is 90° from each other in the azimuthal direction. Conjugate heat transfer between fluid and structure is accounted for. The geometry simplifies a control-rod guide tube (CRGT) in a boiling water reactor (BWR). LES results are compared with measurement data. This is one of the first times BWR conditions are met in both experiments and LES: pressure equals 7.2 MPa, while the temperature difference between hot and cold inlets reaches 216 K. LES temperatures at the fluid-structure interface are fairly correlated with their experimental equivalents, with regard to mean values, local variances, and dangerous oscillation modes in fatigue-prone areas (z = 0.65-0.67 m). An elastic analysis of the structure is performed to evaluate stress intensities there. From them, cumulative fatigue usage factors are estimated and used as screening criteria in the subsequent frequency analysis of temperature time series at the fluid-structure interface. Cracks are likely to initiate after 97 h.

  • 3.
    Bergagio, Mattia
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Thiele, Roman
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Analysis of temperature fluctuations caused by mixing of non-isothermal water streams at elevated pressure2017In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 104, p. 979-992Article in journal (Refereed)
    Abstract [en]

    Temperatures were measured at the inner surface of an annulus between two coaxial tubes, where three water streams mixed. These temperatures were sampled at either 100 Hz or 1000 Hz. The acquisition time was set to 120 s. Two water streams at 549 K, with a Reynolds number between 3.56 × 105 and 7.11 × 105, descended in the annular gap and mixed with a water stream at 333 K or 423 K, with a Reynolds number ranging from 1.27 × 104 to 3.23 × 104. Water pressure was kept at 7.2 MPa. Inner-surface temperatures were collected at eight azimuthal and five axial positions, for each combination of boundary conditions. To better analyze these temperatures and mixing in the vicinity of the wall, scalars estimating the mixing intensity at each measurement position were computed from detrended temperature time series. Fourier and Hilbert–Huang marginal spectra were calculated for the time series giving rise to the highest values of a mixing estimator of choice. The relationship between temperature and velocity was explored by examining the results of an LES simulation using the same boundary conditions as in one of the experimental cases.

  • 4. Gradecka, M.
    et al.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Computational fluid dynamics investigation of supercritical water flow and heat transfer in a rod bundle with grid spacers2016In: Journal of Nuclear Engineering and Radiation Science, ISSN 2332-8983, E-ISSN 2332-8975, Vol. 2, no 3, article id 031015Article in journal (Refereed)
    Abstract [en]

    This paper presents a steady-state computational fluid dynamics approach to supercritical water flow and heat transfer in a rod bundle with grid spacers. The current model was developed using the ANSYS Workbench 15.0 software (CFX solver) and was first applied to supercritical water flow and heat transfer in circular tubes. The predicted wall temperature was in good agreement with the measured data. Next, a similar approach was used to investigate three-dimensional (3D) vertical upward flow of water at supercritical pressure of about 25 MPa in a rod bundle with grid spacers. This work aimed at understanding thermo- and hydrodynamic behavior of fluid flow in a complex geometry at specified boundary conditions. The modeled geometry consisted of a 1.5-m heated section in the rod bundle, a 0.2-m nonheated inlet section, and five grid spacers. The computational mesh was prepared using two cell types. The sections of the rods with spacers were meshed using tetrahedral cells due to the complex geometry of the spacer, whereas sections without spacers were meshed with hexahedral cells resulting in a total of 28 million cells. Three different sets of experimental conditions were investigated in this study: a nonheated case and two heated cases. The nonheated case, A1, is calculated to extract the pressure drop across the rod bundle. For cases B1 and B2, a heat flux is applied on the surface of the rods causing a rise in fluid temperature along the bundle. While the temperature of the fluid increases along with the flow, heat deterioration effects can be present near the heated surface. Outputs from both B cases are temperatures at preselected locations on the rods surfaces. 

  • 5.
    Gradecka, Malwina
    et al.
    Warsaw University of Technology and Oregon State University.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    CFD Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle with Grid Spacers2015In: Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors ISSCWR-7, Helsinki, Finland, 2015Conference paper (Refereed)
    Abstract [en]

    This paper presents steady state CFD simulation approach to supercritical water flow and heat transfer in a rod bundle with grid spacers. The current model was developed using the ANSYS Workbench 15.0 software (CFX solver) and first applied to supercritical water flow and heat transfer in circular tubes. The predicted wall temperature was in good agreement with the measured data. Next, a similar approach was used to investigate three dimensional vertical upward flow of water at supercritical pressure of about 25 MPa in a rod bundle with grid spacers. This work aimed into understanding thermal and hydrodynamic behaviour of fluid flow in complex geometry at specified boundary conditions. The modelled geometry consisted of a 1.5 m heated section in the rod bundle, a 0.2 m non-heated inlet section and five grid spacers. The computational mesh was prepared using two cell types. The sections of the rods with spacers were meshed using tetrahedral cells due to the complex geometry of the spacer, whereas sections without spacers were meshed with hexahedral cells resulting in a total of 28 million cells. Three different sets of experimental conditions were investigated in this study: a non-heated case and two heated cases. The non-heated case, A1, is calculated in order to extract the pressure drop across the rod bundle. For cases B1 and B2 a heat flux is applied on surface of the rods causing a rise in fluid temperature along the bundle. While the temperature of the fluid increases along with the flow heat deterioration effects can be present near the heated surface. Output from both B cases is temperature at pre-selected locations on the rods surfaces. CFD Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle with Grid Spacers.

  • 6. Qvist, Staffan A.
    et al.
    Hellesen, Carl
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Dubberley, Allen E.
    Gradecka, Malwina
    Greenspan, Ehud
    Autonomous Reactivity Control (ARC) - Principles, geometry and design process2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 307, p. 249-274Article in journal (Refereed)
    Abstract [en]

    The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety performance of Generation-IV reactors while having a minimal impact on reactor performance and economic viability. Here we present in detail the principles of how the ARC system operates, what materials should be used, what components make up the system and how they are interconnected. The relevant equations regarding how to design the system for a certain response are developed and defined, and the most important aspects determining the speed of actuation of the systems are analyzed. Thus, this study serves as the general reference material for all of the fundamental principles behind the ARC idea. Finally, we present a step-by-step guide to how a fast reactor fuel subassembly with an ARC system installed would be manufactured, using a full 3D-CAD model. For an ARC installation in a 1000 MWth sodium-cooled oxide fueled fast reactor core, the system constitutes a relatively minor adjustment to a typical fuel assembly, increasing its total axial extent by similar to 5-10% and the total primary coolant pressure drop by similar to 1%. The main finding of this study is that it is possible to design, manufacture (using existing methods) and implement ARC systems in the fuel assemblies of fast reactor cores to provide inherent safety in all anticipated unprotected transients with only a modest increase in the length of the assembly and the pressure drop across the core.

  • 7.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Mechanistic Modeling of Wall-Fluid Thermal Interactions for Innovative Nuclear Systems2015Doctoral thesis, monograph (Other academic)
    Abstract [en]

    Next generation nuclear power plants (GEN-IV) will be capable of not only producing energy in a reliable, safe and sustainable way, but they will also be capable of reducing the amount of nuclear waste, which has been accumulated over the lifetime of current-generation nuclear power plants, through transmutation. Due to the use of new and different coolants, existing computational tools need to be tested, further developed and improved in order to thermal-hydraulically design these power plants.This work covers two different non-unity Prandtl number fluids which are considered as coolants in GEN-IV reactors, liquid lead/lead-bismuth-eutectic and supercritical water. The study investigates different turbulence modeling strategies, such as Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) modeling, and their applicability to these proposed coolants. It is shown that RANS turbulence models are partly capable of predicting wall heat transfer in annular flow configurations. However, improvements in these prediction should be possible through the use of advanced turbulence modeling strategies, such as the use of separate thermal turbulence models. A large blind benchmark study of heat transfer in supercritical water showed that the available turbulence modeling strategies are not capable of predicting deteriorated heat transfer in a 7-rod bundle at supercritical pressures. New models which take into account the strong buoyancy forces and the rapid change of the molecular Prandtl number near the wall occurring during the transition of the fluid through the pseudocritical point need to be developed. One of these strategies to take into account near-wall buoyancy forces is the use of advanced wall functions, which cannot only help in modeling these kind of flows, but also decrease computational time by 1 to 2 orders of magnitude. Different advanced wall function models were implemented in the open-source CFD toolbox OpenFOAM and their performance for different flows in sub- and supercritical conditions were evaluated. Based on those results, the wall function model UMIST-A by Gerasimov is recommended for further investigation and specific modeling tactics are proposed.Near-wall temperature and velocity behavior is important to and influenced by the wall itself. The thermal inertia of the wall influences the temperature in the fluid. However, a more important issue is how temperature fluctuations at the wall can induce thermal fatigue. With the help of LES thermal mixing in a simplified model of a control rod guide tube was investigated, including the temperature field inside the control rod and guide tube walls. The WALE sub-grid turbulence model made it possible to perform LES computations in this complex geometry, because it automatically adapts to near-wall behavior close to the wall, without the use of ad-hoc functions. The results for critical values, such as the amplitude and frequency of the temperature fluctuations at the wall, obtained from the LES computations are in good agreement with experimental results.The knowledge gained from the aforementioned investigations is used to optimize the flow path in a small, passively liquid-metal-cooled pool-type GEN IV reactor, which was designed for training and education purposes, with the help of 3D CFD. The computations were carried out on 1/4 of the full geometry, where the small-detail regions of the heat exchangers and the core were modeled using a porous media approach. It was shown that in order to achieve optimal cooling of the core without changing the global geometry a ratio of close to unity of the pressure drop over the core and the heat exchanger needs to be achieved. This is done by designing a bottom plate which channels enough flow through the core without choking the flow in the core. Improved cooling is also achieved by reducing heat losses from the hot leg through the flow shroud to the cold leg by applying thermal barrier coating similar to methods used in gas turbine design.

  • 8.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Modeling of Direct Contact Condensation With OpenFOAM2010Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    Within the course of the master thesis project, two thermal phase change models for direct contact conden-sation were developed with different modeling approaches, namely interfacial heat transfer and combustionanalysis approach.After understanding the OpenFOAM framework for two phase flow solvers with phase change capabilities,a new solver, including the two developed models for phase change, was implemented under the name ofinterPhaseChangeCondenseTempFoam and analyzed in a series of 18 tests in order to determine the physicalbehavior and robustness of the developed models. The solvers use a volume-of-fluid (VOF) approach withmixed fluid properties.It has been shown that the approach with inter-facial heat transfer shows physical behavior, a strong timestep robustness and good grid convergence properties. The solver can be used as a basis for more advancedsolvers within the phase change class.

  • 9.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Modeling of Direct Contact Condensation With OpenFOAM2010Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    Within the course of the master thesis project, two thermal phase change models for direct contact conden-sation were developed with different modeling approaches, namely inter-facial heat transfer and combustionanalysis approach.

    After understanding the OpenFOAM framework for two phase flow solvers with phase change capabilities, a new solver, including the two developed models for phase change, was implemented under the name of interPhaseChangeCondenseTempFoam and analyzed in a series of 18 tests in order to determine the physical behavior and robustness of the developed models. The solvers use a volume-of-fluid (VOF) approach withmixed fluid properties.

    It has been shown that the approach with inter-facial heat transfer shows physical behavior, a strong timestep robustness and good grid convergence properties. The solver can be used as a basis for more advanced solvers within the phase change class.

  • 10.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Modeling of Direct Contact Condensation With OpenFOAM2010Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
    Abstract [en]

    Within the course of the master thesis project, two thermal phase change models for direct contact conden-sation were developed with different modeling approaches, namely interfacial heat transfer and combustionanalysis approach.After understanding the OpenFOAM framework for two phase flow solvers with phase change capabilities,a new solver, including the two developed models for phase change, was implemented under the name of interPhaseChangeCondenseTempFoam and analyzed in a series of 18 tests in order to determine the physicalbehavior and robustness of the developed models. The solvers use a volume-of-fluid (VOF) approach withmixed fluid properties.It has been shown that the approach with inter-facial heat transfer shows physical behavior, a strong timestep robustness and good grid convergence properties. The solver can be used as a basis for more advancedsolvers within the phase change class.

  • 11.
    Thiele, Roman
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Prediction of forced convection heat transfer to Lead-Bismuth-Eutectic2013Licentiate thesis, monograph (Other academic)
    Abstract [en]

    The goal of this work is to investigate the capabilities of two different commercial codes, OpenFOAM and ANSYS CFX, to predict forced convection heat transfer in low Prandtl number fluids and investigate the sensitivity of these predictions to the type of code and to several input parameters.The goal of the work is accomplished by predicting forced convection heat transfer in two different experimental setups with the codes OpenFOAM and ANSYS CFX using three different turbulence models and varying the input parameters in an extensive sensitivity analysis. The computational results are compared two the experimental data and analyzed for qualitative and quantitative parameters, such as shape of velocity and temperature profiles, thickness of the boundary layers and wall temperatures.The results show that predictions of the temperature and velocity field are generally sufficient to good, however, the sensitivity especially to the turbulent Prandtl number has to be taken into account when computing forced convection heat transfer in low Prandtl number fluids. The results also show that methods applied to OpenFOAM cannot directly be applied to ANSYS CFX.

  • 12.
    Thiele, Roman
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Flow pattern analysis in ELECTRA under natural circulation condition of liquid lead2013In: Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, NURETH-15, Pisa, Italy, 2013Conference paper (Refereed)
    Abstract [en]

    ELECTRA is the European lead cooled training reactor, which is a passively liquid lead cooled reactor of compact size fueled with plutonium nitride. This contribution aims at investigating the flow patterns in the reactor vessel and identifying modifications to the design. The investigation is carried out using 3D computational fluid dynamics. The core and the heat exchanger are modeled using the porous media approach in order to reduce the mesh size. The results show that a CFD approach can be used to model a full system such as ELECTRA. Changes to the thermal hydraulic design to ELECTRA are proposed, which include a redesign of the bottom plate to reduce the bypass flow and this way reduce the temperature difference over the core, which at the current design exceeds the design specifications.

  • 13.
    Thiele, Roman
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Modeling of Forced Convection Heat Transfer to Lead-Bishmuth Eutectic in OpenFOAM2011In: Proceedings of the ANS Wintermeeting 2011, Washington, DC, USA: American Nuclear Society , 2011, p. 1009-1010Conference paper (Refereed)
    Abstract [en]

    In this publication several turbulence models which are implemented in OpenFOAM are validated against experimental data for heat transfer in a lead-bismuth-eutectic flow. The results show that with decreasing Reynolds number the heat transfer predictions degenerate. With lower than standard constant Prandtl number the heat transfer predictions become better. Overall, one can say that the predictions by the k-omega-SST model by Menter (from 2003) give the best predictions, but also need the finest grid (about 2.5 times more cells than comparable models).

  • 14.
    Thiele, Roman
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Numerical modeling of forced-convection heat transfer to lead-bismuth eutectic flowing in vertical annuli2013In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 254, p. 111-119Article in journal (Refereed)
    Abstract [en]

    This paper provides temperature and velocity distribution computations in heated annuli using RANS approach and employing three different turbulent viscosity models. In addition to comparison calculations an extensive sensitivity study was performed. The results show that the RANS approach and the turbulent viscosity models can be used for prediction of forced convection heat transfer to lead–bismuth-eutectic. However, the turbulent Prandtl number has to be carefully selected depending on the respective turbulence model.

  • 15.
    Thiele, Roman
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Optimisation of the Flow Path in a Conceptual Pool Type Reactor under Natural Circulation with Lead Coolant2014In: Proceedings of the NUTHOS-10, Okinawa, Japan: Atomic Society of Japan , 2014Conference paper (Refereed)
    Abstract [en]

    This contribution investigates the eects of a bypass flow blocking bottom plate and the influenceof the heat transfer between the hot and cold leg in a small pool type reactor cooled throughnatural convection with lead coolant. The computations are carried out using 3D computationalfluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled usinga porous media approach. The introduction of full conjugate heat transfer shows that the heattransfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculationzones. These zones become even more pronounced with the introduction of a bottom plate, whichon the other hand also increases the flow through the core and lowers the maximum temperature inthe core by approximately 150 K. Based on the results, redesign suggestions for the bottom plateand the internal wall are made.

  • 16.
    Thiele, Roman
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Investigation of the Influence of Turbulence Models on the Prediction of Heat Transfer to Low Prandtl Number Fluids2011In: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-14, Toronto, ON, Canada: Canadian Nuclear Society , 2011Conference paper (Refereed)
    Abstract [en]

    Despite many advances in computational fluid dynamics (CFD), heat transfer modeling and val-idation of code for liquid metal flows needs to be improved. This contribution aims to providevalidation of several turbulence models implemented in OpenFOAM. 6 different low Reynoldsnumber and 3 high Reynolds number turbulence models have been validated against experimen-tal data for 3 different Reynolds numbers. The results show that most models are able to predictthe temperature profile tendencies and that especially the k-ω-SST by Menter has good predic-tive capabilities. However, all turbulence models show deteriorating capabilities with decreasingReynolds numbers.

1 - 16 of 16
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