| 1. |
- Anglart, Henryk, 1954-
(författare)
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CFD modelling of annular two-phase flow and heat transfer
- 2017
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Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery, Inc.
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Konferensbidrag (refereegranskat)abstract
- 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.
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| 2. |
- Anglart, Henryk, 1954-, et al.
(författare)
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Experimental and numerical investigations of wall temperature fluctuations due to thermal mixing in an annulus
- 2016
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Konferensbidrag (refereegranskat)abstract
- 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.
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| 3. |
- Anglart, Henryk, 1954-
(författare)
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Initial entrained fraction at onset of annular flow
- 2019
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Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 20192019. ; , s. 1023-1034
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Konferensbidrag (refereegranskat)abstract
- One of the frequently employed current models to predict the occurrence of dryout in boiling annular flows is using conservation equations to determine the liquid film mass flux. The accuracy of predictions of dryout depends to a large extend on the initial conditions, which are employed in the model. In this paper it is shown that the accuracy of predictions can be significantly improved if the initial entrained fraction of liquid is correlated to the flow conditions at the onset of annular flow. Using experimental data for liquid film flow rates in pipes with variable power distributions, a new closure relationship for the initial entrained fraction of liquid at the onset of annular flow is proposed.
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| 4. |
- Anglart, Henryk, et al.
(författare)
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Measurement of Wall Temperature Fluctuations during Thermal Mixing of Non-isothermal Water Streams
- 2015
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Ingår i: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). - : American Nuclear Society.
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Konferensbidrag (refereegranskat)abstract
- This paper is dealing with measurement of temperature fluctuations during mixing of two water streams in an annular test section at BWR operational conditions. The experiments are simulating conditions existing in a guide tube of BWR control rods, where relatively cold water at about 333 K is mixing with hot water at ~550 K. It is shown that the mixing is causing high amplitude temperature fluctuations in the solid walls of the control rod extender. Using new movable multi-sensors it became possible to obtain a large experimental database, containing wall temperature measurements at 8 azimuthal and 5 axial positions, with 13 thermocouples at each position. In total 520 temperature readings were performed, each lasting about 2 minutes and recording transient temperature with frequency of at least 100 samples per second and with estimated non-calibrated uncertainty less than 3.9 K. The present experimental results can be used to analyze the governing phenomena during thermal mixing and also to validate CFD conjugate heat transfer models of thermal mixing applied to actual reactor geometries.
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| 5. |
- Anglart, Henryk, 1954-, et al.
(författare)
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Mechanistic modelling of dryout and post-dryout heat transfer
- 2018
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Ingår i: Energy. - : Elsevier. - 0360-5442 .- 1873-6785. ; 161, s. 352-360
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 6. |
- Anglart, Henryk, 1954-
(författare)
-
Progress in understanding and modelling of annular two-phase flows with heat transfer
- 2019
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 345, s. 166-182
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 7. |
- Bergagio, Mattia, et al.
(författare)
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An iterative finite-element algorithm for solving two-dimensional nonlinear inverse heat conduction problems
- 2018
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier. - 0017-9310 .- 1879-2189. ; 126, s. 281-292
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 8. |
- Bergagio, Mattia, et al.
(författare)
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Analysis of temperature fluctuations caused by mixing of non-isothermal water streams at elevated pressure
- 2017
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier. - 0017-9310 .- 1879-2189. ; 104, s. 979-992
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Tidskriftsartikel (refereegranskat)abstract
- Temperatures were measured at the inner surface of an annulus between two coaxial tubes, where three water streams mixed. These temperatures were sampled at either 100 Hz or 1000 Hz. The acquisition time was set to 120 s. Two water streams at 549 K, with a Reynolds number between 3.56 × 105 and 7.11 × 105, descended in the annular gap and mixed with a water stream at 333 K or 423 K, with a Reynolds number ranging from 1.27 × 104 to 3.23 × 104. Water pressure was kept at 7.2 MPa. Inner-surface temperatures were collected at eight azimuthal and five axial positions, for each combination of boundary conditions. To better analyze these temperatures and mixing in the vicinity of the wall, scalars estimating the mixing intensity at each measurement position were computed from detrended temperature time series. Fourier and Hilbert–Huang marginal spectra were calculated for the time series giving rise to the highest values of a mixing estimator of choice. The relationship between temperature and velocity was explored by examining the results of an LES simulation using the same boundary conditions as in one of the experimental cases.
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| 9. |
- Bergagio, Mattia
(författare)
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Experimental analysis of thermal mixing at reactor conditions
- 2016
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- High-cycle thermal fatigue arising from turbulent mixing of non-isothermal flows is a key issue associated with the life management and extension of nuclear power plants. The induced thermal loads and damage are not fully understood yet.With the aim of acquiring extensive data sets for the validation of codes modeling thermal mixing at reactor conditions, thermocouples recorded temperature time series at the inner surface of a vertical annular volume where turbulent mixing occurred. There, a stream at either 333 K or 423 K flowed upwards and mixed with two streams at 549 K. Pressure was set at 72E5 Pa. The annular volume was formed between two coaxial stainless-steel tubes. Since the thermocouples could only cover limited areas of the mixing region, the inner tube to which they were soldered was lifted, lowered, and rotated around its axis, to extend the measurement region both axially and azimuthally.Trends, which stemmed from the variation of the experimental boundary conditions over time, were subtracted from the inner-surface temperature time series collected. An estimator assessing intensity and inhomogeneity of the mixing process in the annulus was also computed. In addition, a frequency analysis of the detrended inner-surface temperature time series was performed. In the cases examined, frequencies between 0.03 Hz and 0.10 Hz were detected in the subregion where mixing inhomogeneity peaked.The uncertainty affecting such measurements was then estimated.Furthermore, a preliminary assessment of the radial heat flux at the inner surface was conducted.
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| 10. |
- Bergagio, Mattia
(författare)
-
Experimental and analytical study of thermal mixing at reactor conditions
- 2018
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- 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.
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| 11. |
- Bergagio, Mattia, et al.
(författare)
-
Experimental investigation of mixing of non-isothermal water streams at BWR operating conditions
- 2017
-
Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 317, s. 158-176
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Tidskriftsartikel (refereegranskat)abstract
- In this experimental investigation, wall surface temperatures have been measured during mixing of three water streams in the annular gap between two coaxial stainless-steel tubes. The inner tube, with an outer diameter of 35 mm and a thickness of 5 mm, holds six K-type, ungrounded thermocouples with a diameter of 0.5 mm, which measured surface temperatures with a sampling rate of either 100 Hz or 1000 Hz. The tube was rotated from 0 to 360° and moved in a range of 387 mm in the axial direction to allow measurements of surface temperatures in the whole mixing region. The outer tube has an inner diameter of 80 mm and a thickness of 10 mm to withstand a water pressure of 9 MPa. A water stream at a temperature of either 333 K or 423 K and a Reynolds number between 1657 and 8410 rose vertically in the annular gap and mixed with two water streams at a temperature of 549 K and a Reynolds number between 3.56E5 and 7.11E5. These two water streams entered the annulus radially on the same axial level, 180° apart. Water pressure was kept at 7.2 MPa. Temperature recordings were performed at five axial and eight azimuthal locations, for each set of boundary conditions. Each recording lasted 120 s to provide reliable data on the variance, intermittency and frequency of the surface temperature time series at hand. Thorough calculations indicate that the uncertainty in the measured temperature is of 1.58 K. The mixing region extends up to 0.2 m downward of the hot inlets. In most cases, measurements indicate non-uniform mixing in the azimuthal direction, because of asymmetries in either geometry or mass flow rates at the hot inlets. Due to the measurement accuracy and a relatively simple geometry, an experimental database has been obtained for validation of computational methods to predict thermal mixing and fatigue. Furthermore, these data can provide new insight into turbulent mixing at BWR operating conditions and, more generally, into mixing coupled to the dynamics, also termed level-2 mixing.
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| 12. |
- Bergagio, Mattia, et al.
(författare)
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Instrumentation for Temperature and Heat Flux Measurement on a Solid Surface under BWR Operating Conditions
- 2015
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Ingår i: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). - : American Nuclear Society. - 9781510811843 ; , s. 5962-5975
-
Konferensbidrag (refereegranskat)abstract
- A new experimental facility has been developed at KTH Royal Institute of Technology to measure temperature and heat flux propagations in solid walls due to mixing of non-isothermal water streams in their vicinity. The main purpose of the measurements has been to obtain a high-precision experimental database suitable for validation of Computational Fluid Dynamics (CFD) codes. Consequently, a set of experiments have been performed in a test section simulating the annular region in the BWR control-rod guide tubes. Since preliminary CFD results implied that 0.1-1 Hz temperature oscillations were to be expected, this experimental research intends to assess the magnitude of temperature fluctuations within the abovementioned frequency range. To this end, water and wall temperatures have been measured in the innermost part of the test-section annulus, with a variety of boundary conditions. As thermocouples would otherwise be available at few axial and azimuthal coordinates only, the tube they are installed on has been lifted, lowered and rotated by a software-controlled motor to record temperature fluctuations in the whole mixing region. At each measurement point, data have been collected over a time long enough to detect the existence of the aforesaid fluctuations. Moreover, an uncertainty analysis has been carried out concerning water temperatures. Thermocouples meant to monitor these temperatures have been modelled with a finite-element method for this very purpose. The wall heat flux has also been estimated using experimental data, thanks to a corrected finite-difference Crank-Nicolson scheme.
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| 13. |
- Fan, Wenyuan, 1991-, et al.
(författare)
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Numerical investigation of spatial and temporal structure of annular flow with disturbance waves
- 2019
-
Ingår i: International Journal of Multiphase Flow. - : Elsevier. - 0301-9322 .- 1879-3533. ; 110, s. 256-272
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 14. |
- Fan, Wenyuan, et al.
(författare)
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Prediction of annular two-phase flow with heat transfer
- 2019
-
Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 5230-5238
-
Konferensbidrag (refereegranskat)abstract
- This paper is presenting recent development of mechanistic models to predict annular two-phase flows and heat transfer. The main interest in simulation of such phenomena results from the need to accurately predict the onset of dryout. In heat exchange devices where the heat flux - rather than the temperature - is controlled (such as, e.g., boiling water reactors) the occurrence of dryout may lead to severe damage of equipment, since the temperature of heated wall can significantly increase and the wall can be melted through. The difficulty in simulation of dryout stems from the fact that it is influenced by phenomena, for which the length scales are ranging from microns to meters. The thickness of the liquid film just before the onset of dryout can be just a few microns, whereas the entrained fraction of liquid is governed by two-phase flow in channels with a few meters in length. To cope with such big characteristic-length span, Computational Fluid Dynamics (CFD) codes are proposed. This approach allows to resolve the phenomena close to the wall and also to account for far-field effects related to the flow history. It is shown that with a proper choice of the level of approximation, and with appropriate closure relationships, CFD codes can be used to predict the occurrence of dryout, taking into account, between others, the effect of disturbance waves, the influence of channel geometry and the presence of flow obstacles.
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| 15. |
- Fan, Wenyuan, et al.
(författare)
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Progress in Phenomenological Modeling of Turbulence Damping around a Two-Phase Interface
- 2019
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Ingår i: Fluids. - : MDPI. - 2311-5521. ; 4:3
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 16. |
- Gradecka, Malwina, 1988-, et al.
(författare)
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CFD Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle with Grid Spacers
- 2015
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Ingår i: Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors ISSCWR-7. - Helsinki, Finland.
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Konferensbidrag (refereegranskat)abstract
- This paper presents steady state CFD simulation approach to supercritical water flow and heat transfer in a rod bundle with grid spacers. The current model was developed using the ANSYS Workbench 15.0 software (CFX solver) and first applied to supercritical water flow and heat transfer in circular tubes. The predicted wall temperature was in good agreement with the measured data. Next, a similar approach was used to investigate three dimensional vertical upward flow of water at supercritical pressure of about 25 MPa in a rod bundle with grid spacers. This work aimed into understanding thermal and hydrodynamic behaviour of fluid flow in complex geometry at specified boundary conditions. The modelled geometry consisted of a 1.5 m heated section in the rod bundle, a 0.2 m non-heated inlet section and five grid spacers. The computational mesh was prepared using two cell types. The sections of the rods with spacers were meshed using tetrahedral cells due to the complex geometry of the spacer, whereas sections without spacers were meshed with hexahedral cells resulting in a total of 28 million cells. Three different sets of experimental conditions were investigated in this study: a non-heated case and two heated cases. The non-heated case, A1, is calculated in order to extract the pressure drop across the rod bundle. For cases B1 and B2 a heat flux is applied on surface of the rods causing a rise in fluid temperature along the bundle. While the temperature of the fluid increases along with the flow heat deterioration effects can be present near the heated surface. Output from both B cases is temperature at pre-selected locations on the rods surfaces. CFD Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle with Grid Spacers.
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| 17. |
- Gradecka, M., et al.
(författare)
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Computational fluid dynamics investigation of supercritical water flow and heat transfer in a rod bundle with grid spacers
- 2016
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Ingår i: Journal of Nuclear Engineering and Radiation Science. - : American Society of Mechanical Engineers (ASME). - 2332-8983 .- 2332-8975. ; 2:3
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 18. |
- Henriksson, Hans, et al.
(författare)
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Long-term operation of the Swedish Centre for Nuclear Technology (SKC): New challenges and solutions in competence building
- 2016
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Ingår i: Proc. European Conf. Nuclear Education and Training (NESTet 2016), Berlin, Germany, May 22-26, 2016.
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Konferensbidrag (refereegranskat)abstract
- The Swedish Centre for Nuclear Technology (Svenskt Kärntekniskt Centrum, SKC) is a national initiative to perform industry relevant research at Swedish universities, and to support dedicated education of direct use to the Swedish nuclear industry.SKC has been the meeting point between industry and academia for 25 years, and has coped with varying needs from industry and political situations. The present situation in the Nordic countries is split: Sweden plans to shut down four out of ten reactors by 2020, while Finland is planning and constructing new reactors. Even without a strong signal to construct new reactors in Sweden, the need for nuclear competence will stay, as we have challenges in front of us to operate and dismantle power plants, operate the intermediate storage facility CLAB in Oskarshamn, and to build and fill the final repositories in Forsmark.The education supported by the SKC at the selected universities will facilitate possible recruitment for nuclear installations. The funding body of SKC consists of all the Swedish Nuclear Power Plants (NPP) situated in Forsmark (three BWRs), Oskarshamn (three BWRs) and Ringhals (one BWR and three PWRs), and the nuclear fuel manufacturer (Westinghouse), while the main research and education is carried out at Chalmers, KTH Royal Institute of Technology and Uppsala University with corresponding in-kind contribution. The research activities cover highly requested studies for today’s nuclear fleet: material embrittlement, stress-corrosion cracking, accident-tolerant fuel development and ageing management for long-term operation (LTO), while the educational part consists of Bachelor and Master programmes as well as elective courses for students outside the main nuclear programmes and contract education.In the master programmes, focus is on e-learning platforms for courses and examination. Examples from such development are to be presented in the full contribution to the conference. Another success story is project based courses in industry, especially within the Bachelor programmes. This is highly appreciated by students, providing a direct contact with future employers, The success of SKC originates from close contact between the funding bodies and academia on many levels: base funding for course preparation, project support in annual calls, and dedicated long-term research funding. But also dedicated industrial experts, and an enthusiasm from academia to enlighten present research issues, as well as strong presence at universities during student fairs and career days. That is LTO of SKC!
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| 19. |
- Li, Haipeng, et al.
(författare)
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CFD model of diabatic annular two-phase flow using the Eulerian-Lagrangian approach
- 2015
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 77, s. 415-424
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Tidskriftsartikel (refereegranskat)abstract
- A computational fluid dynamics (CFD) model of annular two-phase flow with evaporating liquid film has been developed based on the Eulerian-Lagrangian approach, with the objective to predict the dryout occurrence. Due to the fact that the liquid film is sufficiently thin in the diabatic annular flow and at the pre-dryout conditions, it is assumed that the flow in the wall normal direction can be neglected, and the spatial gradients of the dependent variables tangential to the wall are negligible compared to those in the wall normal direction. Subsequently the transport equations of mass, momentum and energy for liquid film are integrated in the wall normal direction to obtain two-dimensional equations, with all the liquid film properties depth-averaged. The liquid film model is coupled to the gas core flow, which currently is represented using the Eulerian-Lagrangian technique. The mass, momentum and energy transfers between the liquid film, gas, and entrained droplets have been taken into account. The resultant unified model for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show favorable agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.
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| 20. |
- Li, Hua, et al.
(författare)
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CFD modeling of annular two-phase flow for dryout prediction
- 2016
-
Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
-
Konferensbidrag (refereegranskat)abstract
- 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 can not 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 engineering applications, such as in boiling water reactors (BWRs). Recent model development on dryout prediction relies on 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 steam-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.
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| 21. |
- Li, Haipeng, et al.
(författare)
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CFD prediction of droplet deposition in steam-water annular flow with flow obstacle effects
- 2017
-
Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 321, s. 173-179
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Tidskriftsartikel (refereegranskat)abstract
- Recent model development on dryout prediction relies on the annular flow modeling, with three fields of gas, droplets and liquid film accounted for. Therefore one unified computational fluid dynamics (CFD) model for annular flow based on Lagrangian particle tracking approach was developed for dryout applications. On the other hand, it is well acknowledged that dryout performance could be improved if flow obstacles are placed in a flow channel. Therefore, to study and predict the dryout, the governing phenomena of droplet deposition and entrainment in annular flow with and without obstacles need to be investigated. The current work tested the CFD model against experimental data from a steam-water flow experiment. Both data with and without obstacles were employed to test the model capability on deposition calculation. The calculated deposition results without obstacles agree reasonably well with both the experimental data and the existing empirical correlations. In case of the flow with obstacles, the calculations also show reasonably good agreement with the experimental data. The current work laid a basis for further work on annular flow model development with dryout capabilities.
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| 22. |
- Li, Haipeng, et al.
(författare)
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Dryout prediction with CFD model of annular two-phase flow
- 2019
-
Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 349, s. 20-26
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 23. |
- Li, Haipeng, et al.
(författare)
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Modeling of annular two-phase flow using a unified CFD approach
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 303, s. 17-24
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Tidskriftsartikel (refereegranskat)abstract
- A mechanistic model of annular flow with evaporating liquid film has been developed using computational fluid dynamics (CFD). The model is employing a separate solver with two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The liquid 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. Both the Eulerian-Eulerian and the Eulerian-Lagrangian approach are used to describe the droplet and vapor motion in the gas core. 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 steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show good agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.
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| 24. |
- Li, Haipeng, et al.
(författare)
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Prediction of dryout and post-dryout heat transfer using a two-phase CFD model
- 2016
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier. - 0017-9310 .- 1879-2189. ; 99, s. 839-850
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Tidskriftsartikel (refereegranskat)abstract
- This study focuses on development of an integrated CFD model for diabatic high quality two-phase flow including tarns-dryout regions from annular-mist regime to mist regime. One unified three-field CFD model accounting for droplets, gas, and liquid film was developed to simulate both pre and post dryout regions, with local models to determine the dryout occurrence. The thin liquid film model was coupled to the gas core flow model, which is described using the Eulerian-Eulerian approach. For the post-dryout region, the various heat and mass transfer mechanisms between the wall, the gas phase, and the droplets were identified, including the wall-gas convective heat transfer, the droplet evaporation, the droplet-wall direct contact heat transfer and the thermal radiation, to calculate the temperature of the wall and the fluid. Of the most interests, dryout location and wall temperature measurements from a post-dryout heat transfer experiment have been used for the validation. Simulation results show that the dramatic temperature excursion could be well captured using current models. Nevertheless, more work will be continued to improve the accuracy of the results.
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| 25. |
- Pegonen, Reijo, et al.
(författare)
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An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code
- 2017
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 311, s. 156-166
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Tidskriftsartikel (refereegranskat)abstract
- The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.
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| 26. |
- Pegonen, Reijo
(författare)
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Development of an Improved Thermal-Hydraulic Modeling of the Jules Horowitz Reactor
- 2017
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at CEA Cadarache research center in France. The reactor will support existing and future nuclear reactor technologies, with the first criticality expected at the end of this decade.The current/reference CEA methodology for simulating the thermalhydraulic behavior of the reactor gives reliable results. The CATHARE2 code simulates the full reactor circuit with a simplified approach for the core. The results of this model are used as boundary conditions in a three-dimensional FLICA4 core simulation. However this procedure needs further improvement and simplification to shorten the computational requirements and give more accurate core level data. The reactor’s high performance (e.g. high neutron fluxes, high power densities) and its design (e.g. narrow flow channels in the core) render the reactor modeling challenging compared to more conventional designs. It is possible via thermal-hydraulic or solely hydraulic Computational Fluid Dynamics (CFD) simulations to achieve a better insight of the flow and thermal aspects of the reactor’s performance. This approach is utilized to assess the initial modeling assumptions and to detect if more accurate modeling is necessary. There were no CFD thermal-hydraulic publications available on the JHR prior to the current PhD thesis project.The improvement process is split into five steps. In the first step, the state-of-the-art CEA methodology for thermal-hydraulic modeling of the reactor using the system code CATHARE2 and the core analysis code FLICA4 is described. In the second and third steps, a CFD thermal-hydraulic simulations of the reactor’s hot fuel element are undertaken with the code STAR-CCM+. Moreover, a conjugate heat transfer analysis is performed for the hot channel. The knowledge of the flow and temperature fields between different channels is important for performing safety analyses and for accurate modeling. In the fourth step, the flow field of the full reactor vessel is investigated by conducting CFD hydraulic simulations in order to identify the mass flow split between the 36 fuel elements and to describe the flow field in the upper and lower plenums. As a side study a thermal-hydraulic calculation, similar to those performed in previous steps is undertaken utilizing the outcome of the hydraulic calculation as an input. The final step culminates by producing an improved, more realistic, purely CATHARE2 based, JHR model, incorporating all the new knowledge acquired from the previous steps.The primary outcome of this four year PhD research project is the improved, more realistic, CATHARE2 model of the JHR with two approaches for the hot fuel element. Furthermore, the project has led to improved thermal-hydraulic knowledge of the complex reactor (including the hot fuel element), with the most prominent findings presented.
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| 27. |
- Pegonen, Reijo, et al.
(författare)
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Hot Fuel Element Thermal-Hydraulic Modeling in the Jules Horowitz Reactor Nominal and LOFA Conditions
- 2015
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Ingår i: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16).
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Konferensbidrag (refereegranskat)abstract
- The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at CEA Cadarache research center in France. The reactor will support the existing and future nuclear reactor technologies and will start operation at the end of the decade.The current CEA methodology for simulating the thermal-hydraulic behavior of the reactor gives reliable results. Today the CATHARE2 code simulates the full reactor with a simplified approach for the core and the boundary conditions are transferred into the three-dimensional FLICA4 core simulation. However this procedure needs to be further improved and simplified to shorten the computational time and to give more accurate core level data. Specific CFD calculations will better identify the thermal-hydraulics phenomena and optimize the meshing/model of the improved procedure.This article presents the current one-coupled thermal-hydraulic modeling of the reactor utilizing the system code CATHARE2 and the core analysis code FLICA4 and describes the more realistic new hot fuel element modeling by using CFD code STAR-CCM+ including conjugate heat transfer. Finally, the results from the both modeling are compared in the hot channel in the nominal condition and in the case of LOFA.This study has improved the thermal-hydraulic knowledge of the complex hot fuel element and the most prominent finds are presented. In addition, the possible improvements for the more realistic CATHARE2’s core model are proposed. In all simulations the safety criteria were satisfied, the reactor stayed in the single-phase regime and overall integrity of the fuel plate was ensured.
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| 28. |
- Pegonen, Reijo, et al.
(författare)
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Hot fuel element thermal-hydraulics in the Jules Horowitz Reactor
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 300, s. 149-160
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Tidskriftsartikel (refereegranskat)abstract
- The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. This paper presents a Computational Fluid Dynamics simulation of the reactors hot fuel element. Moreover conjugate heat transfer analysis is performed for the hot channel. The main objective of this work is to improve the thermal-hydraulic knowledge of the complex hot fuel element and to present the most prominent finds. Possible improvements for the future work are suggested.
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| 29. |
- Pegonen, Reijo, et al.
(författare)
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Hydraulic modeling of the Jules Horowitz Reactor: Mass flow split between 36 fuel elements
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 308, s. 9-19
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Tidskriftsartikel (refereegranskat)abstract
- The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at the CEA Cadarache research center in southern France. The reactor will support existing and future nuclear reactor technologies and the first criticality is expected to be achieved at the end of this decade. This paper presents Computational Fluid Dynamics hydraulic calculations of the reactor and some results of the side thermal-hydraulic simulation of the fuel element. The main objective of this work is to improve the hydraulic knowledge of the reactor and to present the mass flow distribution between 36 fuel assemblies. Potential improvements for future work are proposed.
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| 30. |
- Riber Marklund, Anders, 1982-, et al.
(författare)
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Demonstration of an improved passive acoustic fault detection method on recordings from the Phénix steam generator operating at full power
- 2017
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 101, s. 1-14
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Tidskriftsartikel (refereegranskat)abstract
- A hidden Markov model method proposed earlier for passive acoustic leak detection in sodium fast reactor systems has been improved in order to clarify how to set all free model parameters and to allow smaller amounts of training data. The method is based on training the model on known background noise only and optimizing its free model parameters by a parametric study of detection performance for synthetic noises superposed onto the same background. This means that the method is not assuming any knowledge on the noise to be detected and may be used as a general fault detection method, even if the application envisaged here is leak detection for sodium fast reactors. Using recordings of background noise as well as from argon injection tests performed at full power in the Phénix sodium fast reactor plant, it is estimated that the resulting method will detect leak-like deviations from the background noise with a detection delay of a few seconds, a false alarm rate close to 10-8 per second and at signal-to-noise ratio conditions at least corresponding to an additive signal at −10 dB. The method is one-channel, i.e. using input from one single acoustic sensor only.
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| 31. |
- Riber Marklund, Anders, 1982-
(författare)
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Passive acoustic leak detection in energy conversion systems of sodium fast reactors
- 2016
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Reaching the goals of Generation IV nuclear power is challenging. However, no less than six reactor concepts have been identified as capable of fulfilling the demands. Among these, the Sodium Fast Reactor (SFR), probably represents the most mature technology as about 20 SFR plants have been operated to this day. One design-specific issue of the SFR is the risk of leak and sodium-water reaction inside a steam generator. Standard monitoring is based on hydrogen detection, resulting in high sensitivity but slow response. The alternative of acoustic leak detection methods has been studied since the 1970s since they are able to respond much faster. Demonstrating low false alarm rate while detecting the fairly weak and possibly unknown acoustic signals of leaks has however proven to be difficult. Today, the CEA performs R&D, notably within the scope of the ASTRID project, with the aim of eliminating the sodium-water reaction risk. This is achieved by a Brayton cycle, using a nitrogen turbine and compact sodium-nitrogen heat exchangers. In case of a leak in this system, the low solubility of nitrogen in sodium and the high pressure in the tertiary circuit would increase the secondary pressure, locally deteriorate performance and possibly result in harmful hydrodynamic effects. Together with the risks of a potential gas leak over to the reactor, this motivates the use of leak detection also for this design. This thesis concerns passive acoustic leak detection, primarily for a SFR sodium-nitrogen heat exchanger, arguing that this method is suitable based on experiments, numerical simulations and studies on algorithms. The word “passive” here refers to a system that does not send out any signals, but rather records the noise of the plant and detects leaks as changes in this signal. The thesis covers experiments on normal operation and leak-simulating setups as well as machine-learning based detection methods intended to be of interest also for change detection in general.
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| 32. |
- Spirzewski, M., et al.
(författare)
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An improved phenomenological model of annular two-phase flow with high-accuracy dryout prediction capability
- 2018
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 331, s. 176-185
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Tidskriftsartikel (refereegranskat)abstract
- 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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| 33. |
- Spirzewski, Michal, et al.
(författare)
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Uncertainty and sensitivity analysis of a phenomenological dryout model implemented in DARIA system code
- 2019
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Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 355
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Tidskriftsartikel (refereegranskat)abstract
- Uncertainties of numerical predictions play important role in assessment of safety margins in nuclear reactors. In this paper uncertainties of the predictions of the Hewitt-Govan model, which is used to determine the entrainment and deposition rates in phenomenological dryout model, are presented and discussed. Results of the global uncertainty analysis are shown in terms of trends of uncertainty values with respect to the pressure, mass flux and inlet subcooling. Regions of large and small uncertainties are identified and presented in a tabular form. Application of the global sensitivity analysis methods allowed to quantify the sources of uncertainties with high accuracy over large spectrum of experimental conditions. Since the present analysis required calculation of millions of cases, a dedicated fast-running three-field computational code called DARIA has been developed for that purpose. The analysis revealed that the uncertainties of dryout prediction are predominantly resulting from uncertainties of the coolant mass flow rate. The most obvious implication of the presented work is that, when applying the current methodology to predict dryout in rod bundles, the most important parameter, affecting the accuracy of predictions, is the distribution of coolant flow between sub-channels.
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| 34. |
- Thiele, Roman, 1984-
(författare)
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Mechanistic Modeling of Wall-Fluid Thermal Interactions for Innovative Nuclear Systems
- 2015
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Next generation nuclear power plants (GEN-IV) will be capable of not only producing energy in a reliable, safe and sustainable way, but they will also be capable of reducing the amount of nuclear waste, which has been accumulated over the lifetime of current-generation nuclear power plants, through transmutation. Due to the use of new and different coolants, existing computational tools need to be tested, further developed and improved in order to thermal-hydraulically design these power plants.This work covers two different non-unity Prandtl number fluids which are considered as coolants in GEN-IV reactors, liquid lead/lead-bismuth-eutectic and supercritical water. The study investigates different turbulence modeling strategies, such as Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) modeling, and their applicability to these proposed coolants. It is shown that RANS turbulence models are partly capable of predicting wall heat transfer in annular flow configurations. However, improvements in these prediction should be possible through the use of advanced turbulence modeling strategies, such as the use of separate thermal turbulence models. A large blind benchmark study of heat transfer in supercritical water showed that the available turbulence modeling strategies are not capable of predicting deteriorated heat transfer in a 7-rod bundle at supercritical pressures. New models which take into account the strong buoyancy forces and the rapid change of the molecular Prandtl number near the wall occurring during the transition of the fluid through the pseudocritical point need to be developed. One of these strategies to take into account near-wall buoyancy forces is the use of advanced wall functions, which cannot only help in modeling these kind of flows, but also decrease computational time by 1 to 2 orders of magnitude. Different advanced wall function models were implemented in the open-source CFD toolbox OpenFOAM and their performance for different flows in sub- and supercritical conditions were evaluated. Based on those results, the wall function model UMIST-A by Gerasimov is recommended for further investigation and specific modeling tactics are proposed.Near-wall temperature and velocity behavior is important to and influenced by the wall itself. The thermal inertia of the wall influences the temperature in the fluid. However, a more important issue is how temperature fluctuations at the wall can induce thermal fatigue. With the help of LES thermal mixing in a simplified model of a control rod guide tube was investigated, including the temperature field inside the control rod and guide tube walls. The WALE sub-grid turbulence model made it possible to perform LES computations in this complex geometry, because it automatically adapts to near-wall behavior close to the wall, without the use of ad-hoc functions. The results for critical values, such as the amplitude and frequency of the temperature fluctuations at the wall, obtained from the LES computations are in good agreement with experimental results.The knowledge gained from the aforementioned investigations is used to optimize the flow path in a small, passively liquid-metal-cooled pool-type GEN IV reactor, which was designed for training and education purposes, with the help of 3D CFD. The computations were carried out on 1/4 of the full geometry, where the small-detail regions of the heat exchangers and the core were modeled using a porous media approach. It was shown that in order to achieve optimal cooling of the core without changing the global geometry a ratio of close to unity of the pressure drop over the core and the heat exchanger needs to be achieved. This is done by designing a bottom plate which channels enough flow through the core without choking the flow in the core. Improved cooling is also achieved by reducing heat losses from the hot leg through the flow shroud to the cold leg by applying thermal barrier coating similar to methods used in gas turbine design.
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