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  • 1.
    Anglart, Henryk
    et al.
    KTH, Superseded Departments (pre-2005), Physics. 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.
    Anglart, Henryk
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Li, Haipeng
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Niewinski, Grzegorz
    Warsaw Univ Technol, Inst Heat Engn, Nowowiejska 21-25, PL-00665 Warsaw, Poland..
    Mechanistic modelling of dryout and post-dryout heat transfer2018In: Energy, ISSN 0360-5442, E-ISSN 1873-6785, Vol. 161, p. 352-360Article in journal (Refereed)
    Abstract [en]

    In this paper a new mechanistic model for the diabatic annular two-phase flow is presented and applied to prediction of dryout and post-dryout heat transfer in various channels. The model employs a computational fluid dynamics code - OpenFOAM (R) - to solve the governing equations of two-phase mixture flowing in a heated channel. Additional closure laws have been implemented to calculate the location of the dryout and to predict wall temperature in the post-dryout region. Calculated results have been compared with experimental data obtained in pipes and good agreement between predictions and measurements has been achieved. The presented model is applicable to complex geometries and thus can be used for prediction of post-dryout heat transfer in a wide variety of energy conversion systems.

  • 3.
    Bergagio, Mattia
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Experimental and analytical study of thermal mixing at reactor conditions2018Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    High-cycle thermal fatigue due to turbulent mixing of streams at distinct temperatures is an interdisciplinary issue affecting safety and life extension of existing reactors together with the design of new reactors. It is challenging to model damage and thermal loads arising from the above mixing.

    In order to collect vast data sets for the validation of codes modeling turbulent thermal mixing under reactor conditions, temperatures were sampled at the inner surface of the vertical annular volume between two concentric 316LN stainless steel tubes. This annulus simplifies that between control-rod guide tube and stem in Swedish boiling water reactors (BWRs) Oskarshamn 3 and Forsmark 3. In 2008, several stems there were reported as broken or cracked from thermal fatigue. Cold water entered the annulus at 333 K, at axial level z = 0.15 m. It moved upward and mixed with hot water, which entered the annulus at 549 K, at z = 0.80 m. Pressure read 7.2 MPa. Hot and cold inlet temperatures and pressure match BWR conditions. The thermocouples attached to the inner tube could only acquire inner-surface temperatures at six locations, so the inner tube was translated and rotated about the z-axis to expand the measurement zone.

    Mixing inhomogeneity was estimated from such measurements. In the cases examined, the inner-surface temperatures from areas with the highest mixing inhomogeneity show dominant frequencies lower than ten times the inverse of the experiment time.

    The uncertainty of this temperature measurement appears to be equal to 1.58 K.

    A large eddy simulation (LES) of mixing in the above annulus was conducted. Experimental boundary conditions were applied. The conjugate heat transfer between water and tubes was modeled. The wall-adapting local eddy viscosity (WALE) subgrid model was adopted. A finite element analysis (FEA) of the inner tube was performed using LES pressure and temperature as loads. Cumulative fatigue usage factors (CUFs) were estimated from FEA stress histories. To this end, the rainflow cycle-counting technique was applied. CUFs are highest between z = 0.65 m and z = 0.67 m. Cracking is predicted to initiate after 97 h. LES and experimental inner-surface temperatures agree reasonably well in relation to mean values, ranges, mixing inhomogeneity, and critical oscillation modes in areas sensitive to fatigue. LES inner-surface temperatures from areas with the highest CUFs show dominant frequencies lower than ten times the inverse of the simulation time.

    A robust, effective iterative algorithm for reconstructing the transient temperature field in the inner tube from redundant boundary data was implemented and verified. Temperature-dependent properties were included. Initial conditions and over-specified boundary data in the inverse problem were perturbed with Gaussian noise to check the robustness of the solving method to noise.

  • 4.
    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.

  • 5.
    Bergagio, Mattia
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Li, Haipeng
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Warsaw University of Technology, Poland.
    An iterative finite-element algorithm for solving two-dimensional nonlinear inverse heat conduction problems2018In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 126, p. 281-292Article in journal (Refereed)
    Abstract [en]

    It is often useful to determine temperature and heat flux in multidimensional solid domains of arbitrary shape with inaccessible boundaries. In this study, an effective algorithm for solving boundary inverse heat conduction problems (IHCPs) is implemented: transient temperatures on inaccessible boundaries are estimated from redundant simulated measurements on accessible boundaries. A nonlinear heat equation is considered, where some of the material properties are dependent on temperature. The IHCP is reformulated as an optimization problem. The resulting functional is iteratively minimized using a conjugate gradient method together with an adjoint (dual) problem approach. The associated partial differential equations are solved using the finite-element package FEniCS. Tikhonov regularization is introduced to mitigate the ill-posedness of the IHCP. The accuracy of the implemented algorithm is assessed by comparing the solutions to the IHCP with the correct temperature values, on the inaccessible boundaries. The robustness of our method is tested by adding Gaussian noise to the initial conditions and redundant boundary data in the inverse problem formulation. A mesh independence study is performed.

  • 6.
    Gallego-Marcos, Ignacio
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Steam condensation in a water pool and its effect on thermal stratification and mixing2018Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    The Pressure Suppression Pool (PSP) of a Boiling Water Reactor (BWR) is a large heat sink designed to limit the containment pressure by condensing steam released from the primary coolant system. The development of thermal stratification is a safety concern since it leads to higher containment pressures than in completely mixed conditions, and can affect the performance of systems such as the emergency core cooling and containment spray, which the use PSP as a source of water.The goal of this thesis is to develop and validate models for the prediction of the PSP behavior during a steam injection in a Nordic BWR. The framework of the Effective Heat Source and Effective Momentum Source (EHS/EMS) models is used to provide the integral heat and momentum sources induced by the steam condensation. The EHS/EMS can be implemented in a containment thermal-hydraulic or a CFD code, where the pool is modelled with a single-phase liquid solver.EHS/EMS models are developed for the low steam mass flux regimes appearing in (i) large diameter blowdown pipes connecting the drywell to the wetwell pool; and (ii) multi-hole sparger pipes connecting the primary system to the pool.Empirical correlations are developed to predict the effective momentum induced by chugging in the blowdown pipes. The correlations are implemented in GOTHIC, where a containment model is proposed to enable capturing the feedback between pool conditions and drywell pressure. Validation is performed against the PPOOLEX experiments.Conceptual designs are proposed for a set of large-scale pool experiments with spargers in the PPOOLEX and PANDA facilities. Correlations are proposed for the erosion velocity of a cold layer, and ranges are estimated for the angle, profile and turbulence of the momentum sources created by steam injection. CFD simulations of the experiments is done to calibrate the momentum sources in the oscillatory bubble regimes. A concept of the Separate Effect Facility (SEF) is proposed to provide a measurements of the effective momentum. Empirical correlations for the bubble radius, velocity, heat transfer coefficient, etc. are also developed and compared to available data from the literature.Application of the developed CFD and EHS/EMS models to full-scale containment behavior shows that thermal stratification can occur during prototypic steam injection conditions. Recommendations are given on how to avoid this.

  • 7.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression PoolIn: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
    Abstract [en]

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

  • 8.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    Kudinov, Pavel
    Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression PoolManuscript (preprint) (Other academic)
    Abstract [en]

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

  • 9.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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 experiments2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 337, p. 300-316Article in journal (Refereed)
    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.

  • 10.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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 experimentsManuscript (preprint) (Other academic)
    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.

  • 11.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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 regimeIn: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533Article in journal (Refereed)
    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.

  • 12.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    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 regimeManuscript (preprint) (Other academic)
    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.

  • 13.
    Galushin, Sergey
    et al.
    employer.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Analysis of the Effect of Severe Accident Scenario on the Vessel Lower Head Failure in Nordic BWR using MELCOR code2018Conference paper (Refereed)
  • 14.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Royal Institute of Technology.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Sensitivity Analysis of the Vessel Lower Head Failure in Nordic BWR using MELCOR Code2018Conference paper (Refereed)
  • 15.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. employer.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Risk Analysis Framework for Decision Support for Severe Accident Mitigation Strategy in Nordic BWR2018Conference paper (Refereed)
  • 16.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. employer.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Ranlöf, Lisa
    Lloyd's Register Consulting.
    Bäckström, Ola
    Lloyd's Register Consulting.
    Adolfsson, Yvonne
    Lloyd's Register Consulting.
    Marklund, Anders Riber
    Lloyd's Register Consulting.
    Joint Application of Risk Oriented Accident Analysis Methodology and PSA Level 2 to Severe Accident Issues in Nordic BWR2018Conference paper (Refereed)
  • 17. 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. 

  • 18.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors2018Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

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

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

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

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

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

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

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

  • 19.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed)
  • 20.
    Jeltsov, Marti
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Pre-test analysis of an LBE solidification experiment in TALL-3DIn: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed)
    Abstract [en]

    This paper presents the process of development of a solidification test section design for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop of prototypic height). Coolant solidification is a phenomenon of potential safety importance for liquid metal cooled fast reactors (LMFRs). Solidification, e.g. in case of excessive performance of the passive decay heat removal systems, can affect local heat transfer and even lead to partial or complete blockage of the coolant flow paths. This might lead to failure of decay heat removal function. In case of reduced flow circulation, the temperature of the coolant will increase, which might prevent complete blockage of the flow. Prediction of possible outcomes of such scenarios with complex interactions between local physical phenomena of solidification and system scale natural circulation behavior is subject to modelling (epistemic) uncertainty. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support experiment development. The aim of the experimental design is to satisfy requirements stemming from the process of qualification of the model that can be used for addressing the safety-related concerns. In this work we focus on two aspects: (i) design of solidification test section (STS) for investigation of local solidification phenomena of lead-bismuth eutectic (LBE), and (ii) effect of the STS on the system scale behavior of the experimental facility. We discuss selection of the STS characteristics and experimental test matrix using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.

  • 21.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Application of a system thermal-hydraulics code to development of validation process for coupled STH-CFD codes2018Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    Generation IV reactors are designed to provide sustainable energy generation, minimize waste production and excel in safety. Due to lack of operational experience, ever evolving design and stringent safety requirements, these novel reactors have to rely heavily on simulations.

    Best estimate one-dimensional (1D) system thermal-hydraulics (STH) codes, originally intended for simulating water-cooled reactor systems with high coolant mass flow rates, are unable to capture complex three-dimensional (3D) phenomena in liquid metal cooled pool-type reactors. Computational fluid dynamics (CFD) codes are capable of resolving the 3D effects, however applying these methods with high resolution for the whole primary system results in prohibiting computational cost.

    At the same time, there are system components where flow can, with reasonable accuracy, be approximated with 1D models (e.g. core channels, some heat exchangers, etc.). One of the proposed solutions in order to achieve adequate accuracy and affordable computational efficiency in modelling of a Generation IV reactor is to divide the primary system into 1D and 3D regions and apply coupled STH and CFD codes on the respective sub-domains.

    Successful validation is a prerequisite for application of both, standalone and coupled STH and CFD codes in design and safety analysis of Generation IV systems. In this work we develop and apply different aspects of code validation methodology with an emphasis on (i) STH code analysis in support of validation experiment design (facility and test conditions), (ii) calibration of uncertain code input parameters and validation of standalone STH code, (iii) development of an approach to couple STH and CFD codes.

    A considerable part of the thesis work is related to the development of a loop-type, 3 leg, liquid metal experimental facility TALL-3D for code validation. Particular focus was on identification of test conditions featuring complex feedbacks between 1D-3D phenomena, which can be challenging for the codes. Standalone STH code (RELAP5) was validated against experimental data. The domain of natural circulation instabilities in TALL-3D operation parameters was discovered using a validated STH code and global optimum search algorithms. Then existence of growing natural circulation oscillations was experimentally confirmed. An international benchmark was initiated in the framework of EU SESAME project based on the obtained experimental data.

    Simulations were performed to define dimensions and location of a new test section for coolant solidification experiments that would also enhance possibilities for studying natural circulation instabilities in the future tests.

    An approach to automated input calibration and code validation is developed in order to minimize possible “user effect” in case of multiple uncertain input parameters (UIPs) and system response quantities (SRQs). These methods were applied extensively in the development of RELAP5 input models and identification of the natural circulation instability regions.

    Domain overlapping approach to coupling of RELAP5 and Star-CCM+ codes was proposed and resulted in considerable improvement of the predictive capabilities in comparison to standalone RELAP5.

  • 22.
    Kööp, Kaspar
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Automated calibration and validationof RELAP5 input model against TALL-3D facility experimental dataIn: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed)
    Abstract [en]

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

  • 23.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Optimal neutron population growth in accelerated Monte Carlo criticality calculations2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 117, p. 297-304Article in journal (Refereed)
    Abstract [en]

    We present a source convergence acceleration method for Monte Carlo criticality calculations. The method gradually increases the neutron population size over the successive inactive as well as active criticality cycles. This helps to iterate the fission source faster at the beginning of the simulation where the source may contain large errors coming from the initial cycle; and, as the neutron population size grows over the cycles, the bias in the source gets reduced. Unlike previously suggested acceleration methods that aim at optimisation of the neutron population size, the new method does not have any significant computing overhead, and moreover it can be easily implemented into existing Monte Carlo criticality codes. The effectiveness of the method is demonstrated on a number of PWR full-core criticality calculations using a modified SERPENT 2 code.

  • 24. Plukienė, R.
    et al.
    Plukis, A.
    Juodis, L.
    Remeikis, V.
    Šalkauskas, O.
    Ridikas, D.
    Gudowski, Wacław
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 330, p. 241-249Article in journal (Refereed)
    Abstract [en]

    The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT. The IAEA nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE 6 code package has been used for the determination of the RBMK-1500 spent nuclear fuel composition. The total amount of trans-uranium elements has been estimated in both countries by 2015. Major parameters of the hybrid system performance (e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation cases. Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system. About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the 3 GWth fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition. The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation time has been estimated.

  • 25. Spirzewski, M.
    et al.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Warsaw University of Technology, Poland.
    An improved phenomenological model of annular two-phase flow with high-accuracy dryout prediction capability2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 331, p. 176-185Article in journal (Refereed)
    Abstract [en]

    This paper presents a new phenomenological model of annular two-phase flow with dryout prediction capability, implemented in the CATHARE-3 system code. The model comprises existing correlations for entrainment and deposition rates and a new equation to determine the initial entrained fraction (IEF) of the liquid phase at the onset of annular two-phase flow. The proposed new model allows for a significant reduction of mean error variations with pressure and mass flux, when compared with measured dryout in pipes with internal diameter from 8 to 14.9 mm, system pressure from 3 to 10 MPa, mass flux from 500 to 6000 kg/m2s, test section length from 1 to 7 m, inlet subcooling form 10 to 100 K, and critical heat flux from 0.15 to 3.90 MW/m2. It has been also shown that, at certain conditions, the phenomenological model is unable to provide an accurate prediction, irrespective of the chosen value for the IEF parameter. Such behavior is thoroughly investigated in this paper and seldom addressed in the literature, even though it sets limits on the applicability of the model to dryout predictions.

  • 26.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Albanova University Centre, 10691 Stockholm, Sweden.
    Bortot, Sara
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Albanova University Centre, 10691 Stockholm, Sweden.
    A small lead-cooled reactor with improved Am-burning and non-proliferation characteristics2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 193-200Article in journal (Refereed)
    Abstract [en]

    In this paper, a novel approach for transmutation of americium in fast reactors is presented. Using enriched uranium as fissile support, rather than plutonium, it is shown that a minor actinide burning rate of 25 kg/TWhth is possible to achieve in a passively safe, critical lead-cooled reactor. Moreover, the plutonium produced by transmutation of 241Am features up to 38% 238Pu, making it difficult to use for weapons production. 

  • 27.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Qvist, S.
    LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden..
    Mickus, Ignas
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Bortot, Sara
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Szakalos, Peter
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Surface and Corrosion Science. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Ejenstam, Lina
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Surface and Corrosion Science.
    Design of SEALER, a very small lead-cooled reactor for commercial power production in off-grid applications2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 338, p. 23-33Article in journal (Refereed)
    Abstract [en]

    In this paper, the conceptual design of a small lead-cooled nuclear reactor intended to replace diesel-power in off-grid applications is presented. In a vessel of dimensions making it transportable by air, the targeted design performance is to produce 3 MW of electrical power for up to 30 years without reloading of fuel. Consequently, the inner vessel can be sealed, delaying malevolent access to the nuclear fuel and improving security. Alumina forming alloys are applied to ensure long term corrosion protection of fuel cladding tubes, steam generator tubes and primary vessel over the operational temperature regime. Moreover, decay heat can be removed in a completely passive manner by natural convection from the core to the primary coolant and by thermal radiation from the primary vessel to the environment. Finally, the source term is such that relocation of population residing beyond 1 km from the reactor will not be required even in the case of a complete core melt.

1 - 27 of 27
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