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Prof. Dr. Rui Li

  • Physics
  • Thermodynamics
  • Energy technology
  • Numerical simulation
  • Process and reactor safety

Faculty of European Campus Rottal-Inn

Professors

Professor

AA 222

0991/3615-8857

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Zeitschriftenartikel

  • F. Gabrielli
  • W. Maschek
  • Rui Li
  • C. Matzerath Boccaccini
  • M. Flad
  • S. Gianfelici
  • B. Vezzoni
  • A. Rineiski

Probabilistic evaluation of the energetics upper bound during the transition phase of an unprotected loss of flow accident for a sodium cooled fast reactor by using a Phenomenological Relationship Diagram

In: Nuclear Engineering and Design vol. 341

  • (2019)

DOI: 10.1016/j.nucengdes.2018.11.004

Abstract anzeigen

One of the main research goals of the GEN-IV systems is enhancing their safety compared with the former Sodium-Cooled Fast Reactor (SFR) designs. A key issue is the capability of accidents prevention as well as of demonstrating that their consequences do not violate the safety criteria. In order to fulfill such requirements, risk analyses of severe core disruptive accidents are performed. Since the beginning of the SFR development, Hypothetical Core Disruptive Accidents (HCDAs) have played an outstanding role. Numerous safety analyses have been performed for developing and licensing past SFR designs and nowadays a large database of results is available. In particular, a large amount of results of the mechanistic SIMMER-II and SIMMER-III/IV analyses for various core designs and different power classes is available at the Karlsruhe Institute of Technology (KIT). The current paper describes the probabilistic approach based on the Phenomenological Relationship Diagram (PRD), which is used to evaluate the Probability Distribution Function (PDF) of the thermal energy release during the transition phase of an unprotected loss of flow accident scenario for a SFR. The technique allows taking into account the mechanistic nature of the accident scenario. In fact, the available results of the mechanistic analyses of HCDAs in SFRs are used to assess the PDFs of the dominant phenomena affecting the thermal energy release, which are propagated in the PRD by employing a Monte Carlo method.
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Zeitschriftenartikel

  • W. Maschek
  • Rui Li
  • C. Matzerath Boccaccini
  • F. Gabrielli
  • K. Morita

Investigation on upper bounds of recriticality energetics of hypothetical core disruptive accidents in sodium cooled fast reactors

In: Nuclear Engineering and Design vol. 326 pg. 392-402.

  • (2018)

DOI: 10.1016/j.nucengdes.2017.11.002

Abstract anzeigen

One key research goal for GEN-IV systems is an enhanced safety compared to the former Sodium Cooled Fast Reactor concepts. A key issue is built-in safety and the capability to prevent accidents and to demonstrate that their consequences do not violate aimed-at safety criteria. From the beginning of SFR development the Core Disruptive Accident (CDA) has played an outstanding role in the safety assessment. Under core disruptive accident conditions with core melting the fuel might compact, prompt criticality might be achieved and a severe nuclear power excursion with mechanical energy release might be the consequence. Numerous safety analyses accompanied the development and the licensing procedures of past fast reactor projects. A central issue of all analyses was the assessment of a realistic upper bound of energetics especially related to recriticalities in disrupted core configurations. Striving for an even higher safety level for next generation reactors a new strategy focused on the development and introduction of preventive and mitigative measures both to reduce the chance for a severe accident development and to mitigate its energetics. For assessing the effectiveness of these measures the knowledge of the CDA behavior is essential. In this context and on basis of new code developments, new experimental insights and extended studies for many reactor types of different power classes over the recent years, the issue of a realistic upper bound of energetics of the late core melt phases is again of relevance. Of special interest is the identification of natural and intrinsic mechanisms that limit the escalation of energetics. The current paper deals with these issues and tries to add supportive facts on the limits of CDA energetics. The evaluation of results of mechanistic SIMMER-II and SIMMER-III/IV analyses performed for various core designs and power classes and specific model case studies in 2D and 3D geometry indeed supports the idea of a limit of recriticality energetics. Intrinsic mechanisms exist, which limit the escalation energetics even in case of a strong blockage confinement suppressing any fuel discharge and allowing on-going sloshing recriticalities. In the light of the available information and taking into account relevant scientific publications and studies by the international community on the subject, one could conclude that an upper bound for energetics in the range given in the paper can be deduced.
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Zeitschriftenartikel

  • Rui Li
  • X.-N. Chen
  • L. Andiolo
  • A. Rineiski

3D numerical study of LBE-cooled fuel assembly in MYRRHA using SIMMER-IV code

In: Annals of Nuclear Energy vol. 104 pg. 42-52.

  • (2017)

DOI: 10.1016/j.anucene.2017.02.009

Abstract anzeigen

The present paper is based on the work carried out in the framework of the European FP7 project MAXSIMA, in which MYRRHA safety studies are performed. MYRRHA is a pool-type 100 MWth system with MOX fuel designed to operate both in ADS and critical modes. It uses lead-bismuth eutectic (LBE) as primary coolant. The MOX fuel has almost the same density as the LBE coolant. In case of pin failure fuel pellets may break into chunks and particles carried by the coolant upwards and redistributed in the reactor pool. The transmutation group at IKET/KIT mainly with the numerical analysis tool is involved for studying severe accidents for MYRRHA reactor design. The highlight of the current work is that 3D simulations with explicit modelling on the gaps between fuel assemblies and 3D macroscopic pin bundle models are performed for the first time using a reactor safety analysis code, SIMMER-IV, with 3D geometry. In this paper, the numerical analyses are conducted for a single fuel assembly blockage and 19 pin-rods on basis of an LBE coolant experiment. The 3D analysis has been applied with both scales namely fuel assembly scale and pin bundle scale. For the fuel assembly scale, the evaluation of a single fuel assembly blockage using non-axisymmetric geometry configuration in the subcritical mode is addressed. For the pin bundle scale, the 3D pin bundle simulations show a good agreement from the experiment conducted at KALLA liquid metal laboratory. Note that this is the only code applied by now for blockage analyses after pin failure in MYRRHA. The current work has formed a solid basis for the safety analysis for MYRRHA in the future.
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Zeitschriftenartikel

  • X.-N. Chen
  • Rui Li
  • F. Belloni
  • F. Gabrielli
  • A. Rineiski
  • L. Andriolo
  • L. Guo
  • D. Castelliti
  • M. Schyns
  • E. Bubelis
  • G. Bandini
  • M. Sarotto

Safety studies for the MYRRHA critical core with the SIMMER-III code

In: Annals of Nuclear Energy vol. 110 pg. 1030-1042.

  • (2017)

DOI: 10.1016/j.anucene.2017.08.021

Abstract anzeigen

The presented studies are carried out within the European 7th framework project MAXSIMA, in which the MYRRHA reactor, which stands for Multi-purpose hYbrid Research Reactor for High-tech Applications, developed at SCK-CEN (Belgian Nuclear Research Centre), is investigated. The SIMMER code is employed for severe accident investigations of the reactor at KIT and SCK-CEN in both critical and ADS subcritical modes. In this paper only studies for the critical core are presented. The SIMMER-III model has been set up and assessed first for the neutronic feedback coefficients. Its calculated fuel, coolant and structure feedbacks agree well with the results evaluated by means of the European Reactor ANalysis Optimized System (ERANOS). For benchmarking of the SIMMER-III coupled neutronics and fluid-dynamics model, several Unprotected Transients due to Over Power (UTOP) have been calculated and compared with results of transient system codes. Very good agreement is demonstrated. In case of the largest and quickest reactivity insertion under hypothetical accident conditions, the reactor is assumed to turn for a short time into a slightly prompt supercritical state, but a quite mild power excursion takes place. Blockage accidents are studied in detail with SIMMER only. In total three scenarios have been investigated, namely the blockages of a single fuel assembly (FA), the protected core blockage scenario and the damage propagation of defective pin failures. Our studies demonstrated no core damage propagation can possibly occur under the different blockage scenarios.
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Zeitschriftenartikel

  • Rui Li
  • W. Maschek
  • C. Matzerath Boccaccini
  • B. Vezzoni
  • M. Flad
  • A. Rineiski

Impact of the Bell–Plesset instability on centralized sloshing in pool geometry

In: International Journal of Hydrogen Energy vol. 41 pg. 7126-7131.

  • (2016)

DOI: 10.1016/j.ijhydene.2016.01.152

Abstract anzeigen

The theoretical and numerical analyses have been conducted to investigate the kinetic energy attenuation characteristics on basis of identical geometry and liquid properties together with an existing centralized sloshing experiment. The goal of this paper is to assess the quantitative impact of the Bell–Plesset (BP) instability on the sloshing motion. The results show that BP instability plays a certain role in azimuthal energy dissipation when the sloshing waves are moving inwards and converging in cylindrical geometry. The velocity attenuation is calculated via a perturbation flow equation assuming the initial perturbation length 1 mm, it shows that the velocity could be suppressed by 25% due to BP instability. The corresponding simulation using particle method has been performed. With the help of numerical simulation, the initial perturbation length is approximated as 1.3 mm which is in line with the assumption in the theoretical analysis.
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Zeitschriftenartikel

  • Rui Li
  • W. Maschek
  • C. Matzerath Boccaccini
  • M. Marchetti
  • V. Kriventsev
  • A. Rineiski

Bell-Plesset Instability Analysis for an Inward Centralized Sloshing

In: Nuclear Engineering and Design vol. 297 pg. 312-319.

  • (2016)

DOI: 10.1016/j.nucengdes.2015.12.010

Abstract anzeigen

Liquid sloshing is a typical phenomenon when liquid in a container has an unrestrained surface. In fast reactors under core disruptive accidents (CDAs) conditions specific sloshing motions could be encountered that can be described as a centralized sloshing. It is important to investigate the mitigating and augmenting factors for such centralized sloshing motions. Any retardation or instability effects that reduce the compaction speed and resulting reactivity ramp rate are of importance, requiring an understanding of the kinetic energy dissipation of an inward centralized slosh. In this paper, the Bell–Plesset (BP) instability has been studied theoretically and numerically based on a corresponding inward centralized sloshing experiment. The theoretical analysis is based on the classical perturbation theory and the simulation has been conducted by a fully mesh-free, Lagrangian particle numerical method. With our experimental data, the initial perturbation length 1.3 mm is approximated by the numerical calculation as supplement of the purely theoretical analysis. The outward and inward sloshing timings have been re-checked from the experiment that the inward velocity is reduced by around 20% compared to outward velocity. It experimentally confirms reasonably well the numerical result 17.5%. The experimental, numerical and theoretical analysis show that BP instability plays a certain role in azimuthal energy dissipation when the sloshing waves are moving inwards and converging in cylindrical geometry, for the experiment case the velocity reduction may be 17.5%.
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Zeitschriftenartikel

  • L. Guo
  • Rui Li
  • S. Wang
  • M. Flad
  • W. Maschek
  • A. Rineiski

Numerical investigation of SIMMER code for fuel-coolant interaction

In: International Journal of Hydrogen Energy vol. 41 pg. 7227-7232.

  • (2016)

DOI: 10.1016/j.ijhydene.2016.01.080

Abstract anzeigen

Fuel-coolant interaction (FCI) is a very complex but important issue in the safety analysis of the severe accidents for nuclear reactors due to the rapid multiple thermos–hydrodynamic activities. Until now, there are still large uncertainties existing in various phases during the FCI process, such as the melt solidification, fragmentation and relocation, film boiling on the melt surface, coolant vaporization and following vapor explosion, and so on. SIMMER-III code was first developed to analyses core disruptive accidents in liquid-metal fast reactors (LMFRs) as an integral numerical tool coupling multiphase thermal hydraulic code with neutron kinetics model, and was demonstrated its reasonable flexibility in some FCI simulations. In this paper, the applicability of the code in simulating the premixing phase of FCI process is verified in comparison with a related jet-type experiment in literature. In addition, the sensitivity analysis on several key parameters of the related models in the SIMMER code was performed to assess the impacts in the simulation of the FCI premix phase. It is expected that the results can provide some numerical experience for the uncertainty analysis of FCI calculation using SIMMER-III code.
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Zeitschriftenartikel

  • X.-N. Chen
  • Rui Li
  • A. Rineiski
  • W. Jäger

Macroscopic pin bundle model and its blockage simulations

In: Energy Conversion and Management vol. 91 pg. 93-100.

  • (2015)

DOI: 10.1016/j.enconman.2014.11.053

Abstract anzeigen

In this paper a macroscopic continuum differential model of pin bundle flow is proposed and developed for computational fluid dynamics (CFD) simulations of a reactor core. Thereby the pin bundle flow is regarded as a porous medium flow that is characterized by a certain coolant volume fraction and an associated wet area. The frictional drags experienced by pins and wrappers in the axial and radial directions are converted to pressure drops, i.e. momentum exchange terms, which are therefore anisotropic from the macroscopic point of view. Such a model reduces the number of CFD meshes very much and can be applied for a whole reactor core flow simulation without losing details of subchannel flow. In particular the model is implemented in the SIMMER-III code and applied for the MYRRHA reactor design. A steady state of subchannel flow, which is considerably non-uniform in the radial direction, is investigated and compared with a subchannel code. Satisfactory agreements are achieved. As a practical example the subchannel blockage in the central channels is considered and simulated. The scenario of pin failure and fuel sweep-out is expected, but it can take place already at 50% area blockage in a fuel assembly, if the blockage is located at the entrance of the active zone.
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Zeitschriftenartikel

  • Rui Li
  • X.-N. Chen
  • C. Matzerath Boccaccini
  • A. Rineiski
  • W. Maschek

Study on Severe Accident Scenarios: Pin Failure Possibility of MYRRHA-FASTEF Critical Core

In: Energy Procedia vol. 71 pg. 14-21.

  • (2015)

DOI: 10.1016/j.egypro.2014.11.850

Abstract anzeigen

The present work is carried out within the European FP7 project SEARCH, in which the MYRRHA demonstrator reactor is designed to be able to operate both in ADS mode and in critical mode using lead-bismuth eutectic (LBE) as primary coolant. According to the project task definition, the pin failure and fuel dispersion scenarios in severe accidents have to be extensively studied for reactor safety analysis. In this paper, the unprotected severe transients analyses for the MYRRHA-FASTEF critical core were performed using SIMMER-III code. The aim of the current work was to obtain a deeper understanding of core material redistribution processes before and after pin damage, since the Archimedes force could move pellets, chunks and fuel particles upwards out of the core and redistribute them into the upper pool region and peripheral structures. Starting the simulations with the steady state calculation, relevant parameters reflect good agreement with the design operational conditions. For the transients three postulated severe accident scenarios were proposed that may possibly lead to pin failure and furthermore core damage: unprotected loss of flow (ULOF), unprotected transient overpower (UTOP) and unprotected blockage accident (UBA), where in particular the entrance of fuel assembly is blocked as a side window is still open. The three transients, starting from the steady state conditions, have been investigated. The calculation results show for the MYRRHA-FASTEF that under the conditions chosen all simulated transient cases do not lead to a pin failure and fuel redistribution.
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Zeitschriftenartikel

  • Rui Li
  • X.-N. Chen
  • A. Rineiski
  • D. Zhang
  • E. Merle-Lucotte

Transient analyses for a molten salt fast reactor with optimized core geometry

In: Nuclear Engineering and Design vol. 292 pg. 164-176.

  • (2015)

DOI: 10.1016/j.nucengdes.2015.06.011

Abstract anzeigen

Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.
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Zeitschriftenartikel

  • Rui Li
  • X.-N. Chen
  • A. Rineiski
  • V. Moreau

Studies of fuel dispersion after pin failure: Analysis of assumed blockage accidents for the MYRRHA–FASTEF critical core

In: Annals of Nuclear Energy vol. 79 pg. 31-42.

  • (2015)

DOI: 10.1016/j.anucene.2015.01.002

Abstract anzeigen

The present work has been carried out in the framework of the European FP7 project SEARCH, in which the MYRRHA demonstrator reactor is designed to be able to operate both in ADS mode and in critical mode using lead–bismuth eutectic (LBE) as primary coolant. According to the project task definition, the pin failure and fuel dispersion scenarios in severe accidents had to be extensively studied for reactor safety analysis. In this paper, the unprotected severe transients analyses for the MYRRHA–FASTEF critical core were performed using the SIMMER-III code. The aim of the current work is to obtain a deeper understanding of core material redistribution processes after pin damage. Since the fuel has almost the same density as the coolant, its pellets, chunks and particles will essentially be carried by the coolant flow, thus moving upwards out of the core and redistributing into the upper pool region and peripheral structures. Starting the simulations from the steady state configuration, relevant parameters reflect good agreement with the design operational conditions. For the transients, the most severe accident scenario proposed, that may possibly lead to pin failure and furthermore core damage, is the unprotected blockage accident (UBA). The calculation results show that after pin failure, the mobile fuel starts to re-distribute. In the meantime, the reactor stabilizes to shut-down status because of the fuel loss. Our results show that the blockage propagation is impossible thanks to the gap between fuel subassemblies.
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Zeitschriftenartikel

  • Rui Li
  • M. Mori
  • H. Ninokata

A calculation methodology proposed for liquid droplet impingement erosion

In: Nuclear Engineering and Design vol. 242 pg. 157-163.

  • (2012)

DOI: 10.1016/j.nucengdes.2011.10.004

Abstract anzeigen

Bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in nuclear reactors. Liquid droplet impingement (LDI) erosion could be regarded as one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plant safety. In this paper a computational methodology is established for simulation of LDI erosion using computational fluid dynamics (CFD) simulation and theoretical calculation. Two-phase flow numerical simulations are conducted for standard elbow geometry, typically with the pipe diameter of 170 mm. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ɛ turbulence model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The turbulence damping in vapor–droplets flow is theoretically analyzed by a damping function on the energy spectrum basis of single phase flow. Locally, a droplet impact angle function is employed to determine the overall erosion rate. Finally, the overall and local investigations are combined to purpose a general methodology of LDI erosion prediction procedure, which has been complemented into CFD code. Based on our more physical computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena.
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Zeitschriftenartikel

  • Rui Li
  • M. Pellegrini
  • H. Ninokata
  • M. Mori

A numerical study on turbulence attenuation model for liquid droplet impingement erosion

In: Annals of Nuclear Energy vol. 38 pg. 1279-1287.

  • (2011)

DOI: 10.1016/j.anucene.2011.02.010

Abstract anzeigen

The bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in the nuclear reactors. The liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants safety. In this paper two-phase numerical simulations are conducted for standard elbow geometry, typically the pipe diameter is 170 mm. The turbulence attenuation in vapor-droplets flow is analysed by a damping function on the energy spectrum basis of single phase flow. Considering the vapor turbulent kinetic energy attenuation due to the involved droplets, a computational fluid dynamic (CFD) tool has been adopted by using two-way vapor-droplet coupled system. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ε model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach, a general LDI erosion prediction procedure for bent pipe geometry has been performed to supplement the CFD code. The liquid droplets diameter, velocity, volume concentration are evaluated for the effects of carrier turbulence attenuation. The result shows that carrier turbulence kinetic energy attenuation is proved to be an important effect for LDI erosion rate when investigating the bent pipe wall thinning phenomena.
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Zeitschriftenartikel

  • Rui Li
  • H. Ninokata
  • M. Mori

A numerical study of impact force caused by liquid droplet impingement onto a rigid wall

In: Progress in Nuclear Energy vol. 53 pg. 881-885.

  • (2011)

DOI: 10.1016/j.pnucene.2011.03.002

Abstract anzeigen

Liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the inner bent pipe surface. Evaluating the LDI erosion is an important topic of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants. One of the causes of LDI erosion is the impact pressure by the impingement of droplets in the involved steam. We investigated a simple droplet impingement to a rigid wall using volume of fluid (VOF) model, which is a two-phase Eulerian–Eulerian approach. The impact of a single water droplet with a high velocity towards a solid surface is examined numerically. The high Reynolds number value implies inertia dominated the phenomena and supports an inviscid approach to the problem. The high Weber number is justifying that an assumption to neglect the surface tension effect is adopted. We show that the compressibility of the liquid medium plays a dominant role in the evolution of the phenomenon. Both generation and propagation of shock waves are directly computed by solving the fluid dynamics continuity and momentum equations. In the simulation we employed a front tracking solution procedure, which is particularly suitable for two-phase free surface computation. The numerical results show that critical maximum pressure is not highest at the center of droplet contact on the surface at the first instantaneous moment but highest behind the contact angle later before jet eruption. It agrees generally well (within 20%) with the mathematical analysis. Finally, a droplet impact angle function is proposed for the global LDI erosion prediction.
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Zeitschriftenartikel

  • Rui Li
  • A. Yamaguchi
  • H. Ninokata

Computational Fluid Dynamics Study of Liquid Droplet Impingement Erosion in the Inner Wall of a Bent Pipe

In: Journal of Power and Energy Systems vol. 4 pg. 327-336.

  • (2010)

DOI: 10.1299/jpes.4.327

Abstract anzeigen

The bent pipe wall thinning phenomenon has been often found at the elbow of pipelines in the power engineering industry. Liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the inner bent pipe surface. In this paper, three-dimensional numerical simulations are conducted for a bent pipe. Typically the pipe diameter is 170mm and the bending angle is 90 degree, the mass flow rate of droplet is 4.5×10-3 kg/s with the velocity of 280m/s at the entry. The calculations employ a two-phase flow model. A computational fluid dynamic tool has been adopted by using one-way and two-way fluid-droplet coupled system in high Reynolds number regions. This computational fluid model is built up by incompressible Reynolds averaged Navier-Stokes equations using different turbulent flow computational models and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The momentum transfers between droplet and carrier fluid are calculated by using two different fluid-droplet coupled methods. The interactional force between carrier and droplet are taken into account by momentum transfer in Eulerian-Lagrangian approaches. Based on the carrier streamlines and droplet trajectories, the two-way calculation using the interactional momentum transfer calculations could be a more appropriate model to simulate the bent pipe wall thinning phenomena, the effects of droplet size are also demonstrated numerically. Finally, it is shown that turbulence models are not sensitive to the involved droplets.
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Zeitschriftenartikel

  • Rui Li
  • E. Merzari
  • H. Ninokata

Numerical Study on Liquid Droplet Impingement Erosion in BWRs

In: Transactions of the American Nuclear Society vol. 101 pg. 863-864.

  • (2009)
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Zeitschriftenartikel

  • Rui Li
  • Y.-L. He
  • Y.-G. Lei
  • Y. Tao
  • P. Chu

A Numerical Study on Fluid Flow and Heat Transfer Performance of Internally Roughened Tubes with Dimples

In: Journal of Enhanced Heat Transfer vol. 16 pg. 267-285.

  • (2009)

DOI: 10.1615/JEnhHeatTransf.v16.i3.40

Abstract anzeigen

In this paper, 3D numerical simulations are conducted in heat transfer enhanced tubes with internally roughened dimples in the range of Re = 2000 to 11,000. The fluid flow and heat transfer characteristics are fully understood by the local and overall friction factors, Nusselt number, and Colburn j factor. The local mean friction factor has a periodic change due to the periodic dimple distribution, the eddy zone and high-pressure region of every dimple greatly influence the working fluid for heat transfer enhancement. In this numerical simulation, five samples are employed to study the detailed heat transfer characteristics. The results show that the dimpled tube is a new kind of heat transfer enhanced tube with the excellent heat transfer performance and low resistance. It is found that the effects of dimples on the heat transfer performance can be well described by the field synergy principle. The effect of different dimple arrangements is very little, which is always within 2%. But the effect of dimple size on heat transfer and fluid flow performance is very significant, thus there exists a tube with an optimum dimple size among the tubes investigated in this paper. By integrated performance evaluation of NUF0/NU0F, a maximum of about 60% heat transfer enhancement with the same friction penalty can be gained by the optimal dimpled tube.
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Zeitschriftenartikel

  • Y.-G. Lei
  • Y.-L. He
  • Rui Li
  • Y.-F. Gao

Effects of baffle inclination angle on flow and heat transfer of a heat exchanger with helical baffles

In: Chemical Engineering and Processing: Process Intensification vol. 47 pg. 2336-2345.

  • (2008)

DOI: 10.1016/j.cep.2008.01.012

Abstract anzeigen

Numerical simulations were carried out to study the impacts of various baffle inclination angles on fluid flow and heat transfer of heat exchangers with helical baffles. The simulations were conducted for one period of seven baffle inclination angles by using periodic boundaries. Predicted flow patterns from simulation results indicate that continual helical baffles can reduce or even eliminate dead regions in the shell side of shell-and-tube heat exchangers. The average Nusselt number increases with the increase of the baffle inclination angle α when α < 30°. Whereas, the average Nusselt number decreases with the increase of the baffle inclination angle when α > 30°. The pressure drop varies drastically with baffle inclination angle and shell-side Reynolds number. The variation of the pressure drop is relatively large for small inclination angle. However, for α > 40°, the effect of α on pressure drop is very small. Compared to the segmental heat exchangers, the heat exchangers with continual helical baffles have higher heat transfer coefficients to the same pressure drop. Within the Reynolds number studied for the shell side, the optimal baffle inclination angle is about 45°, with which the integrated heat transfer and pressure drop performance is the best. The detailed knowledge on the heat transfer and flow distribution in this investigation provides the basis for further optimization of shell-and-tube heat exchangers.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG

Vita

  • 2018 - today: Professor “Energy Technology - Gas and Heat Networks”, Deggendorf Institute of Technology
  • 2012 - 2018: Researcher, Institute for Thermal Energy Technology and Safety (former: Institute for Nuclear and Energy Technology) Faculty of Mechanical Engineering, Karlsruhe Institute of Technology (KIT)
  • 2008 - 2012: PhD student, Laboratory for Advanced Nuclear Energy Faculty of Mechanical Engineering, Tokyo Institute of Technology