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  • 1.
    Adorno Lopes, Denise
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Wilson, T. L.
    Univ South Carolina, Columbia, SC 29208 USA..
    Kocevski, V.
    Univ South Carolina, Columbia, SC 29208 USA..
    Moore, E. E.
    Univ South Carolina, Columbia, SC 29208 USA..
    Besmann, T. M.
    Univ South Carolina, Columbia, SC 29208 USA..
    Wood, E. Sooby
    Univ Texas San Antonio, San Antonio, TX USA..
    White, J. T.
    Los Alamos Natl Lab, Los Alamos, NM USA..
    Nelson, A. T.
    Los Alamos Natl Lab, Los Alamos, NM USA..
    Middleburgh, S. C.
    Westinghouse Elect Sweden AB, Vasteras, Sweden.;Bangor Univ, Nucl Futures Inst, Bangor LL57 1UT, Gwynedd, Wales..
    Claisse, Antoine
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Experimental and computational assessment of U-Si-N ternary phases2019In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 516, p. 194-201Article in journal (Refereed)
    Abstract [en]

    Uranium nitride-silicide composites are being considered as a high-density and high thermal conductivity fuel option for light water reactors. During development, chemical interactions were observed near the silicide melting point which resulted in formation of an unknown U-Si-N ternary phase. In the present work, U-Si-N composite samples were produced by arc-melting U3Si2 under an argon-nitrogen atmosphere to form the ternary phase. The resulting samples were characterized by SEM/EDS-EPMA and XRD, and demonstrated an equilibrium between U3Si2, UN, USi and a U-Si-N phase with a distinct crystallographic structure. Rietveld refinement of the ternary structure was performed, considering the ternary structures existent in the analogue U-Si-C system, and a good fit was obtained for the hexagonal U(20)Si(16)N(3 )phase. DFT + U calculations were performed in parallel to evaluate the thermodynamic and dynamic stability of the ternaries U20Si16N3 and U3Si2N2. The calculated enthalpy of formation and phonon dispersion support the existence of stable U20Si16N3 and U3Si2N2, although some soft modes in the U(20)Si(16)N(3)( )phase phonons are observed. The results presented here thus demonstrate the occurrence of at least one ternary phase in the U-Si-N system.

  • 2.
    Anglart, Henryk
    KTH, School of Industrial Engineering and Management (ITM), Energy Technology. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    CFD modelling of annular two-phase flow and heat transfer2017In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Conference paper (Refereed)
    Abstract [en]

    This paper describes the governing phenomena and current approaches in their modeling for annular two-phase flow and heat transfer. The complexity of the flow, including liquid film, disturbance waves, turbulent gas core, droplet deposition and entrainment, are discussed. Computational Fluid Dynamics (CFD) approach to model the phenomena is presented. 

  • 3.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Progress in understanding and modelling of annular two-phase flows with heat transfer2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 345, p. 166-182Article in journal (Refereed)
    Abstract [en]

    Annular two-phase flows with heat transfer play important role in many industrial applications. In particular, thermal margins of Boiling Water Reactors (BWR) are entirely determined by this type of flow and heat transfer conditions. To avoid dryout, a liquid film must be present on heated rods of BWR fuel assemblies during normal operation. The present paper describes the recent progress in understanding and modelling of the governing phenomena of annular two-phase flow and heat transfer. A special attention has been devoted to experimental observations that have the most significant influence on the adopted modelling approach. The primary goal is to pave a path to mechanistic modelling of dryout and post-dryout heat transfer applicable to nuclear fuel assemblies. Current Computational Fluid Dynamics (CFD) approaches to model the governing phenomena are presented and their further improvements are suggested.

  • 4.
    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.

  • 5.
    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.

  • 6.
    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.

  • 7.
    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.

  • 8.
    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.

  • 9. Bonny, G.
    et al.
    Domain, C.
    Castin, N.
    Olsson, Pär
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Malerba, L.
    The impact of alloying elements on the precipitation stability and kinetics in iron based alloys: An atomistic study2019In: Computational materials science, ISSN 0927-0256, E-ISSN 1879-0801, Vol. 161, p. 309-320Article in journal (Refereed)
    Abstract [en]

    Iron based industrial steels typically contain a large number of alloying elements, even so-called low alloyed steels. Under irradiation, these alloying elements form clusters that have a detrimental impact of the mechanical properties of the steel. The stability and formation mechanisms of such clusters are presently not fully understood. Therefore, in this work, we study the thermal stability and formation kinetics of small solute clusters in the bcc Fe matrix. We use density functional theory (DFT) to characterize the binding energy of vacancy/solute clusters containing Cr, Mn, Ni, Cu, Si and P, thereby exploring >700 different configurations. The constructed DFT data base is used to fit a cluster expansion (CE) for the vacancy-FeCrMnNiCuSiP system. In turn, the obtained CE is applied in atomistic kinetic Monte Carlo simulations to study the effect of Mn, Ni, Cr, Si and P on the precipitation formation in the FeCu alloy. We conclude that the addition of Mn and Ni delay the precipitation kinetics by an order of magnitude. The additional alloying with traces of P/Si further delays the kinetics by an additional order of magnitude. We found that Si plays an essential role in the formation of spatially mixed MnNiCuSi cluster

  • 10.
    Christopher, Petersson
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Erosion-Corrosion experiments on Steels in liquid lead and Development of Slow Strain Rate testing rig2019Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
  • 11. Ekberg, Christian
    et al.
    Ribeiro Costa, Diogo
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Hedberg, Marcus
    Jolkkonen, Mikael
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Nitride fuel for Gen IV nuclear power systems2018In: Journal of Radioanalytical and Nuclear Chemistry, ISSN 0236-5731, E-ISSN 1588-2780, Vol. 318, p. 1713-1725Article in journal (Refereed)
    Abstract [en]

    Nuclear energy has been a part of the energy mix in many countries for decades. Today in principle all power producing reactors use the same techniqe. Either PWR or BWR fuelled with oxide fuels. This choice of fuel is not self evident and today there are suggestions to change to fuels which may be safer and more economical and also used in e.g. Gen IV nuclear power systems. One such fuel type is the nitrides. The nitrides have a better thermal conductivity than the oxides and a similar melting point and are thus have larger safety margins to melting during operation. In addition they are between 30 and 40% more dense with respect to fissile material. Drawbacks include instability with respect to water and a sometimes complicated fabrication route. The former is not really an issue with Gen IV systems but for use in the present fleet. In this paper we discuss both production and recycling potential of nitride fuels.

  • 12.
    Estévez-Albuja, S.
    et al.
    Universidad Politécnica de Madrid (UPM), Spain.
    Gallego-Marcos, Ignacio
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jiménez, G.
    Universidad Politécnica de Madrid (UPM), Spain.
    Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.12020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 136, article id 107027Article in journal (Refereed)
    Abstract [en]

    Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.

  • 13.
    Fan, Wenyuan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. Warsaw Univ Technol, Inst Heat Engn, 21-25 Nowowiejska St, PL-00665 Warsaw, Poland..
    Progress in Phenomenological Modeling of Turbulence Damping around a Two-Phase Interface2019In: FLUIDS, ISSN 2311-5521, Vol. 4, no 3, article id 136Article in journal (Refereed)
    Abstract [en]

    The presence of a moving interface in two-phase flows challenges the accurate computational fluid dynamics (CFD) modeling, especially when the flow is turbulent. For such flows, single-phase-based turbulence models are usually used for the turbulence modeling together with certain modifications including the turbulence damping around the interface. Due to the insufficient understanding of the damping mechanism, the phenomenological modeling approach is always used. Egorov's model is the most widely-used turbulence damping model due to its simple formulation and implementation. However, the original Egorov model suffers from the mesh size dependency issue and uses a questionable symmetric treatment for both liquid and gas phases. By introducing more physics, this paper introduces a new length scale for Egorov's model, making it independent of mesh sizes in the tangential direction of the interface. An asymmetric treatment is also developed, which leads to more physical predictions for both the turbulent kinetic energy and the velocity field.

  • 14.
    Fan, Wenyuan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    varRhoTurbVOF: A new set of volume of fluid solvers for turbulent isothermal multiphase flows in OpenFOAM2019In: Computer Physics Communications, ISSN 0010-4655, E-ISSN 1879-2944, article id 106876Article in journal (Refereed)
    Abstract [en]

    The volume of fluid (VOF) method is a popular approach for multiphase flow modeling. The open-source computational fluid dynamics (CFD) software, OpenFOAM, implements a variety of VOF-based solvers and provides users a wide range of turbulence models. Since isothermal multiphase flows under the VOF framework belong to the variable-density incompressible flow category, the isothermal VOF-based solvers in OpenFOAM fail to use the correct turbulence models. varRhoTurbVOF is designed to solve this issue and with the hope to replace all the corresponding existing solvers in the future. With the object-oriented paradigm, varRhoTurbVOF guarantees the usability, reusability and maintainability of the codes. Aside from turbulence modeling, all other features in the original solvers are preserved in varRhoTurbVOF. Program summary: Program Title: varRhoTurbVOF Program Files doi: http://dx.doi.org/10.17632/4t8z8vzyvs.1 Licensing provisions: GPLv3 Programming language: C++ Supplementary material: http://dx.doi.org/10.17632/7mp25kyb4p.4 Nature of problem: Under the VOF framework, the flow of the isothermal mixture belongs to the variable-density incompressible flow category. For such flows, VOF-based solvers of OpenFOAM fail to construct the correct governing equations for turbulence modeling. varRhoTurbVOF contains a set of newly designed VOF-based solvers which could use the desired governing equations for turbulence quantities. Solution method: varRhoTurbVOF creates a new class for variable-density incompressible turbulence models, which allows reusing the existing turbulence model template classes. A set of VOF-based solvers are then created to be able to construct variable-density incompressible turbulence models.

  • 15.
    Fan, Wenyuan
    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 Univ Technol, Inst Heat Engn, 21-25 Nowowiejska St, PL-00665 Warsaw, Poland..
    Numerical investigation of spatial and temporal structure of annular flow with disturbance waves2019In: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533, Vol. 110, p. 256-272Article in journal (Refereed)
    Abstract [en]

    Droplet entrainment is a crucial process for annular flow in terms of heat and mass transfer. Disturbance wave is believed to be a fundamental phenomenon which is closely related to entrainment. A 3D numerical simulation on disturbance waves and entrainment is carried out by using volume of fluid (VOF) method where no periodic boundary condition is used. Since VOF tracks the interface implicitly, a systematic method is developed for post-processing, with which disturbance waves. ripples, base film, and entrainment process are clearly visualized, and the stochastic and chaotic nature of two-phase flow is confirmed. Surfacewise distributions are generated for main wave parameters, and the streamwise developments of such quantities are shown to be consistent with experimental observations. Predictions for main wave parameters are in reasonable agreement with the experiment and empirical correlations. Current work shows the capability and promising application of investigating disturbance waves and entrainment with VOF method.

  • 16.
    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.

  • 17.
    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.

  • 18.
    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.

  • 19.
    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 pool2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 132, p. 442-450Article 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 with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.

  • 20.
    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.

  • 21.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kapulla, Ralf
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paranjape, Sidharth
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Paladino, Domenico
    Paul Scherrer Inst, Div Nucl Energy & Safety Res, Villigen, Switzerland..
    Laine, Jani
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Puustinen, Markku
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Rasanen, Antti
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Pyy, Lauri
    Lappeenranta Univ Technol, Unit Nucl Safety Res, Lappeenranta, Finland..
    Kotro, Eetu
    Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 347, p. 67-85Article in journal (Refereed)
    Abstract [en]

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

  • 22.
    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.

  • 23.
    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.

  • 24.
    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.

  • 25.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Development of Risk Oriented Accident Analysis Methodology for Assessment of Effectiveness of Severe Accident Management Strategy in Nordic BWR2019Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability as a severe accident management strategy (SAM). In case of a severe accident, the debris ejected from the vessel are expected to fragment, quench and form a debris bed, which is coolable by a natural circulation of water. Success of the existing SAM strategy depends on melt release conditions from the vessel which determine (i) properties of ejected debris and, thus, ex-vessel debris bed coolability, and (ii) potential for energetic melt-coolant interactions (steam explosion). The strategy involves complex interactions between physical phenomena (deterministic) and transient accident scenarios (probabilistic).The aim of this work is further extension, implementation and application of the Risk-Oriented Accident Analysis Methodology (ROAAM) to assessment of the severe accident management strategy effectiveness. ROAAM was originally developed for rare, high-consequence hazards, where both aleatory (stochastic) and epistemic (modeling) uncertainties play a significant role in the risk assessment. The main purpose of ROAAM is to provide the input material to an underlying decision making regarding current safety design acceptance, procedures and possible design modifications.This work reports results of (i) development and implementation of probabilistic framework (ROAAM+) for streamlining sensitivity analysis, uncertainty quantification and risk analysis; (ii) analysis of in-vessel phase of accident progression and melt release conditions in Nordic BWR reactor design with MELCOR code; (iii) analysis of the effect of melt release conditions predicted by MELCOR code on the risk of ex-vessel steam explosion.In ROAAM+, “full models”, such as MELCOR code, are used to develop computationally efficient “surrogate models” to enable extensive uncertainty quantification and failure domain analysis. ROAAM+ analysis identified specific assumptions in MELCOR models, which are currently the major contributors to the uncertainty in the assessment of the SAM effectiveness.

  • 26.
    Galushin, Sergey
    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.
    Implementation of Probabilistic Framework of Risk Analysis Framework for Assessment of Severe Accident Management Effectiveness in Nordic BWRIn: Annals of Nuclear EnergyArticle in journal (Refereed)
  • 27.
    Galushin, Sergey
    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.
    Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code2019Conference paper (Refereed)
  • 28.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR2019Conference paper (Refereed)
  • 29.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR2019In: Nuclear Engineering and Design, Vol. 350, p. 243-258Article in journal (Refereed)
    Abstract [en]

    Nordic Boiling Water Reactors (BWRs) rely on the flooding of the lower drywell (LDW) as a severe accident management (SAM) strategy. The termination of a SA is achieved by fragmenting and quenching of the melt released from the vessel. Success of SAM strategy depends on melt release and water pool conditions. The characteristics of the melt release are the major source of uncertainty in quantification of the risk of SAM failure. Vessel failure and melt release modes are subject to aleatory and epistemic uncertainties at the in-vessel accident progression stage. In this work we focus on predicting the properties of debris relocated to the lower plenum using MELCOR code. We address the effect of epistemic uncertainty in modeling parameters and models in the MELCOR code in different severe accident scenarios on main characteristics of the in-vessel accident progression in Nordic BWRs. Sensitivity analysis is performed to rank the importance of MELCOR modelling parameters and the effect of different MELCOR models is addressed by using different versions of the code. The results provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on code predictions.

  • 30.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code2019In: Science and Technology of Nuclear InstallationsArticle in journal (Refereed)
  • 31.
    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 Code2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, 2018Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, corium is poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties of the debris bed and thus if the bed is coolable or not, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. It is currently recognized that the time and the mode of vessel failure, melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in ROAAM+ Framework. The properties of relocated debris, time and the mode of vessel failure and melt release conditions, including in-vessel/ex-vessel pressure, lower drywell pool depth and temperature, are subject to aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we perform sensitivity analysis for a set of representative cases, to evaluate the effect of MELCOR modelling parameters on the process of core degradation and relocation, and vessel failure mode. Major contributors to the uncertainty in the timing of the vessel failure and amount of melt available for release at the time of failure are identified and discussed in detail. 

  • 32.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 135, article id 106976Article in journal (Refereed)
    Abstract [en]

    Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

  • 33.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 135, article id 106976Article in journal (Refereed)
    Abstract [en]

    Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

  • 34.
    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 BWR2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, 2018Conference paper (Refereed)
    Abstract [en]

    Severe accident management (SAM) in Nordic boiling water reactors (BWRs) employ ex-vessel debris cooling in a deep water pool. The success of the strategy requires formation of a coolable porous debris bed; no energetic steam explosion that can threaten containment integrity. Both scenario (aleatory) and modeling (epistemic) uncertainties are important in the assessment of the failure risks. A consistent approach is necessary for the decision making on whether the strategy is sufficiently effective, or modification of the SAM is necessary. Risk Oriented Accident Analysis Methodology (ROAAM+) is a tool for assessment of failure probability to enable robust decision making, insensitive to remaining uncertainty. The challenge for a decision maker is to distinguish the cases when collecting more knowledge and reduction of uncertainty in risk assessment, or modification of risk management strategy would be the most adequate approach given the safety goals and criteria. When either decision is made, ROAAM+ can provide data for selection of the most efficient implementation of the decision by selecting research priorities or modifying design elements that contribute most to the risk. In this work we discuss different approaches for communication of ROAAM+ framework analysis results and decision support. We focus on connection and integration of ROAAM+ results into risk-informed decision making models used in nuclear industry. The results of risk analysis are used in order to provide necessary insights on conditions when suggested changes in the safety design can be justified, taking into account different aspects of risk.

  • 35.
    Galushin, Sergey
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Ranlöf, L.
    Bäckström, O.
    Adolfsson, Y.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Marklund, A. R.
    Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR2018In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper (Refereed)
    Abstract [en]

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

  • 36. 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. 

  • 37.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 63-75Article in journal (Refereed)
    Abstract [en]

    Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel to fragment and quench core melt and provide long term cooling of the debris. One of the risks associated with this strategy is early containment failure due to ex-vessel steam explosion. Assessment of the risk of steam explosion is subject to significant (i) epistemic uncertainties in modelling and (ii) aleatory uncertainties in scenarios of melt release. For quantification of the uncertainties and the risk a full model (FM) based on TEXAS-V code and a computationally efficient surrogate model (SM) have been previously developed. FM is used to provide a database of solutions that is used for development of a SM, while SM is used in extensive sensitivity and uncertainty analysis. In this work, we compare the risk of containment failure with non-reinforced and reinforced hatch door for metallic and oxidic melt release scenarios. We quantify the error of SM in the approximation of the FM and assess the effect of the approximation uncertainty on risk assessment. We analyze the results and suggest a simplified approach for decision making considering predicted failure probabilities, expected costs, and scenario frequencies.

  • 38.
    Ignas, Mickus
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Response Matrix Reloaded: for Monte Carlo Simulations in Reactor Physics2019Licentiate thesis, comprehensive summary (Other academic)
    Abstract [en]

    This thesis investigates Monte Carlo methods applied to criticality and time-dependent problems in reactor physics. Due to their accuracy and flexibility, Monte Carlo methods are considered as a “gold standard” in reactor physics calculations. However, the benefits come at a significant computing cost. Despite the continuous rise in easily accessible computing power, a brute-force Monte Carlo calculation of some problems is still beyond the reach of routine reactor physics analyses. The two papers on which this thesis is based try to address the computing cost issue, by proposing methods for performing Monte Carlo reactor physics calculations more efficiently. The first method addresses the efficiency of the widely-used k-eigenvalue Monte Carlo criticality calculations. It suggests, that the calculation efficiency can be increased through a gradual increase of the neutron population size simulated during each criticality cycle, and proposes a way to determine the optimal neutron population size. The second method addresses the application of Monte Carlo calculations to reactor transient problems. While reactor transient calculations can, in principle, be performed using only Monte Carlo methods, such calculations take multiple thousands of CPU hours for calculating several seconds of a transient. The proposed method offers a middle-ground approach, using a hybrid stochastic-deterministic scheme based on the response matrix formalism. Previously, the response matrix formalism was mainly considered for steady-state problems, with limited application to time-dependent problems. This thesis proposes a novel way of using information from Monte Carlo criticality calculations for solving time-dependent problems via the response matrix.

     

  • 39.
    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.

  • 40.
    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)
  • 41.
    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.

  • 42.
    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.

  • 43.
    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.

  • 44.
    Li, Haipeng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Dryout prediction with CFD model of annular two-phase flow2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 349, p. 20-26Article in journal (Refereed)
    Abstract [en]

    Two-phase flow and heat transfer are of interest to industrial applications due to its high efficiency. In a diabatic annular two-phase flow, the liquid film is depleted by both entrainment of liquid droplets and by evaporation. When the liquid film experiences almost complete depletion and cannot cover the wall, the heat transfer between the fluid and the channel wall significantly deteriorates, leading to the onset of boiling transition called dryout. While the dryout is milder than the departure from nucleate boiling (DNB) occurring in low quality two-phase flows, it could still challenge and damage the channel wall. As a result, the dryout occurrence needs to accurately predicted and avoided in practice, such as in boiling water reactors (BWRs). Research interests haven been recently focused on dryout prediction with annular flow modeling, with three fields of gas, droplets and liquid film accounted for. In the current study, one unified computational fluid dynamics (CFD) model for annular flow was developed for dryout applications. The model is employing a separate solver of two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The film model is coupled to a solver of three-dimensional conservation equations describing the gas core, which is assumed to contain a saturated mixture of vapor and liquid droplets. All the major interaction phenomena between the liquid film and the gas core flow have been accounted for, including the liquid film evaporation as well as the droplet deposition and entrainment. The resultant unified framework for annular flow has been applied to the swam-water flow with conditions typical for a BWR. The simulation results for the liquid film flow and dryout occurrence show favorable agreements with the available experimental data.

  • 45.
    Maier, Annika Carolin
    et al.
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Applied Physical Chemistry.
    Benarosch, Anna
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    El Jamal, Ghada
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Applied Physical Chemistry.
    Jonsson, Mats
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Applied Physical Chemistry.
    Radiation induced dissolution of U3Si2 - A potential accident tolerant fuel2019In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 517, p. 263-267Article in journal (Refereed)
    Abstract [en]

    To assess the integrity of the accident tolerant fuel candidate U 3 Si 2 under geological repository conditions, the kinetics of γ-radiation- and H 2 O 2 - induced oxidative dissolution was studied. The experiments were performed in aqueous solutions containing 10 mM HCO 3 − and in solutions without added HCO 3 − . The same experiments were also performed on UO 2 for comparison. All experiments were performed using powder suspensions. The experiments show that U 3 Si 2 is less than one order of magnitude more reactive towards H 2 O 2 than is UO 2 . The dissolution yield of U 3 Si 2 slightly exceeds the theoretical yield (23%). In experiments with consecutive additions of H 2 O 2 in HCO 3 − solutions, the reactivity remains constant implying that no significant amount of a secondary phase is formed on the U 3 Si 2 surface. The dissolution of Si closely follows that of U in HCO 3 − solution. In solutions without added HCO 3 − the reactivity towards H 2 O 2 is reduced by a factor less than 2. The dissolution is slightly slower than in HCO 3 − containing solutions but precipitation of U is observed after some time. The results of consecutive additions of H 2 O 2 to the HCO 3 − free system shows that the reactivity is decreasing for every addition. This indicates that a secondary phase is formed. XRD shows that the secondary phase is studtite. The irradiation experiments show that the surface area normalized radiation chemical yields for uranium dissolution for U 3 Si 2 and UO 2 in HCO 3 − solution differ by a factor 5–10 in favour of UO 2 . This difference can largely be attributed to the difference in dissolution yield.

  • 46.
    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.

  • 47.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Roberts, Jeremy A.
    Department of Mechanical and Nuclear Engineering, Kansas State University.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Stochastic-deterministic response matrix method for reactor transients2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, article id 107103Article in journal (Refereed)
    Abstract [en]

    Presented is a stochastic-deterministic, response matrix method for transient analyses of nuclear systems. The method is based on the response matrix formalism, which describes a system by a set of response functions. We propose an approach in which these response functions are computed during a set of Monte Carlo criticality calculations and are later used to formulate a deterministic set of equations for solving a space-time dependent problem. Application of the response matrix formalism results in a set of loosely connected equations, which leads to a favourable linear scaling of the problem. The method offers a simplified approach compared to previously proposed response matrix methods by avoiding phase-space expansions in sets of basis functions. We describe the method starting with the fundamental neutron transport considerations, provide a demonstration on two absorber movement transients in a 3 × 3 assembly PWR mini-core geometry, and compare the solutions against time-dependent Monte Carlo simulations.

  • 48.
    Petrache, C. M.
    et al.
    Univ Paris Saclay, CNRS, IN2P3, Ctr Sci Nucl & Sci Matiere, Batiment 104-108, F-91405 Orsay, France..
    Frauendorf, S.
    Univ Notre Dame, Dept Phys, Indianapolis, IN 46557 USA..
    Lv, B. F.
    Univ Paris Saclay, CNRS, IN2P3, Ctr Sci Nucl & Sci Matiere, Batiment 104-108, F-91405 Orsay, France..
    Astier, A.
    Univ Paris Saclay, CNRS, IN2P3, Ctr Sci Nucl & Sci Matiere, Batiment 104-108, F-91405 Orsay, France..
    Dupont, E.
    Univ Paris Saclay, CNRS, IN2P3, Ctr Sci Nucl & Sci Matiere, Batiment 104-108, F-91405 Orsay, France..
    Guo, S.
    Chinese Acad Sci, Inst Modern Phys, Lanzhou 730000, Gansu, Peoples R China..
    Liu, M. L.
    Chinese Acad Sci, Inst Modern Phys, Lanzhou 730000, Gansu, Peoples R China..
    Zhou, X. H.
    Chinese Acad Sci, Inst Modern Phys, Lanzhou 730000, Gansu, Peoples R China..
    Wang, K. L.
    Chinese Acad Sci, Inst Modern Phys, Lanzhou 730000, Gansu, Peoples R China..
    Greenlees, P. T.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Badran, H.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Cox, D. M.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland.;Lund Inst Technol, Dept Math Phys, S-22362 Lund, Sweden..
    Grahn, T.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Julin, R.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Juutinen, S.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Konki, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland.;CERN, CH-1211 Geneva 23, Switzerland..
    Pakarinen, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Papadakis, P.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland.;Univ Liverpool, Oliver Lodge Lab, Liverpool L69 7ZE, Merseyside, England..
    Partanen, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Rahkila, P.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Sandzelius, M.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Saren, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Scholey, C.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Sorri, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland.;Univ Oulu, Sodankyla Geophys Observ, FIN-99600 Sodankyla, Finland..
    Stolze, S.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland.;Argonne Natl Lab, Div Phys, Argonne, IL 60439 USA..
    Uusitalo, J.
    Univ Jyvaskyla, Dept Phys, POB 35, FI-40014 Jyvaskyla, Finland..
    Cederwall, Bo
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Physics.
    Aktas, Özge
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Physics.
    Ertoprak, Aysegül
    KTH, School of Engineering Sciences (SCI), Physics.
    Liu, Huan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kuti, I
    Hungarian Acad Sci, Inst Nucl Res, H-4001 Debrecen, Hungary..
    Timar, J.
    Hungarian Acad Sci, Inst Nucl Res, H-4001 Debrecen, Hungary..
    Tucholski, A.
    Univ Warsaw, Heavy Ion Lab, Pasteura 5a, PL-02093 Warsaw, Poland..
    Srebrny, J.
    Univ Warsaw, Heavy Ion Lab, Pasteura 5a, PL-02093 Warsaw, Poland..
    Andreoiu, C.
    Simon Fraser Univ, Dept Chem, Burnaby, BC V5A 1S6, Canada..
    Collective rotation of an oblate nucleus at very high spin2019In: Physical Review C: Covering Nuclear Physics, ISSN 2469-9985, E-ISSN 2469-9993, Vol. 99, no 4, article id 041301Article in journal (Refereed)
    Abstract [en]

    A sequence of nine almost equidistant quadrupole transitions is observed in Nd-137. The sequence represents an extremely regular rotational band that extends to a spin of about 75/2 and an excitation energy of approximate to 4.5 MeV above yrast. Cranked mean-field calculations of the Nilsson-Strutinsky type suggest an oblate shape for the band. They reproduce the observed I(I +1) dependence of the rotational energy whereas predicting a pronounced decrease in the deformation, which is the hallmark of antimagnetic rotation.

  • 49. 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.

  • 50. 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.

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