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Prof. Dr. Rui Li

  • Physik, Thermodynamik
  • Numerische Simulation
  • Energietechnik, Reaktorsicherheit
  • Quantum Computing

Fakultät Angewandte Informatik

Professor:innen

Professor

Organiser of “Intel Quantum Computing Challenge”: https://www.th-deg.de/hpc-symposium

AA 222

0991/3615-8857

Sortierung:
Zeitschriftenartikel

  • Helana Lutfi
  • Rui Li
  • Thomas Spittler
  • Sascha Kreiskott
  • Katerina Volchek

Increasing Efficiency in Virtual Teaching in an International Context: E-learning and Instructional Approaches at ECRI

In: Bavarian Journal of Applied Science (Special Issue 2022) vol. 6 pg. 211-225.

  • (2022)

DOI: 10.25929/bjas202291

  • Europan Campus Rottal-Inn
  • DIGITAL
Vortrag

  • T. Shinde
  • Helena Liebelt
  • Rui Li

Preliminary Lattice Boltzmann Method Simulation using Intel® Quantum SDK

In: 21st International Workshop on Advanced Computing and Analysis Techniques in Physics Research (ACAT 2022)

CERN - European Organization for Nuclear Research Bari, Italy

  • 27.10.2022 (2022)

The present work is based on the research within the framework of cooperation between Intel Labs and Deggendorf Institute of Technology, since the Intel® Quantum SDK (Software Development Kit) has recently released. Transport phenomena e.g. heat transfer and mass transfer are nowadays the most challenging unsolved problems in computational physics due to the inherent nature of fluid complexity. As the revolutionary technology, quantum computing opens a grand new perspective for numerical simulation including the computational fluid dynamics (CFD). It is true that the current CFD algorithms based on the different scales (e.g. macroscopic or microscopic) need to be translated into quantum system. In the current work the quantum algorithms have been preliminarily implemented for fluid dynamics using the Intel Quantum SDK, one mesoscopic approach has been applied i.e. to solve the lattice Boltzmann equation. Taking the simplest transport phenomena as a starting point, the preliminary quantum simulation results have been validated with the analytical solution and the classical numerical simulation. The potential of quantum in simulating fluid will be discussed.
  • Angewandte Informatik
  • Hochschulleitung und -einrichtungen
  • DIGITAL
Vortrag

  • Rui Li

Keynote Speech - Quantum computing: a grand era for simulating fluid

In: 21st International Workshop on Advanced Computing and Analysis Techniques in Physics Research (ACAT 2022)

CERN - European Organization for Nuclear Research Bari, Italy

  • 23.-28.10.2022 (2022)

  • Hochschulleitung und -einrichtungen
  • DIGITAL
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

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

Quantitative Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium Cooled Fast Reactor

In: Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17). Proceedings of an International Conference organized by the International Atomic Energy Agency, hosted by the Government of the Russian Federation through the State Atomic Energy Corporation "ROSATOM". null (Proceedings series)

Vienna, Austria

  • (2018)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • D. Castelliti
  • M. Sarotto
  • A. Rineiski
  • A. Ferrari
  • S. Mueller
  • G. Bandini
  • M. Polidorio
  • E. Bubelis
  • W. Jaeger
  • T. Hamidouche
  • F. Belloni
  • X.-N. Chen
  • Rui Li
  • I. Pasichnyk

FP7-MAXSIMA Work Package 2 "Safety Analysis in Support of MYRRHA" Main Outcome and Conclusions

Red Hook, NY

  • (2017)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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. 104C pg. 42-52.

  • (2017)

DOI: 10.1016/j.anucene.2017.02.009

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

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

Probabilistic Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium-Cooled SMR by using a Phenomenological Relationship Diagram

In: Energy Procedia vol. 131 pg. 222-229.

  • (2017)

DOI: 10.1016/j.egypro.2017.09.471

An innovative probabilistic approach based on the Phenomenological Relations Diagram is developed to perform risk analyses of Hypothetical Core Disruptive Accidents in Sodium Fast Reactors (SFRs). The novel method is applied to evaluate the probability distribution of the thermal-to-mechanical energy conversion ratio in an ULOF/Post-Disassembly Expansion Phase scenario in a small- to medium- size SFR. The results provide a comprehensive picture of the fuel energy distribution in the system.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

  • X.-N. Chen
  • A. Rineiski
  • F. Gabrielli
  • L. Andriolo
  • B. Vezzoni
  • Rui Li
  • W. Maschek
  • E. Kiefhaber

Fuel-steel mixing and radial mesh effects in power excursion simulations

In: Annals of Nuclear Energy vol. 90 pg. 26-31.

  • (2016)

DOI: 10.1016/j.anucene.2015.11.041

This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • M. Flad
  • Rui Li
  • S. Wang
  • B. Vezzoni
  • L. Andriolo
  • M. Hartig
  • A. Rineiski
  • W. Maschek

Analyses of CDA of SFR. ESNII+ workshop on severe accident analyses

  • 03.-04.11.2016 (2016)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

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

Application of a Probabilistic Relationship Diagram for PDE Mechanical Energy Release Evaluation after HCDA in a Sodium-Cooled SMR

Sun Valley, ID, USA

  • (2016)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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%.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

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

Recent MYRRHA Safety Studies with the SIMMER Code at KIT

  • (2016)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

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

Probabilistic Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium-Cooled SMR by using a Phenomenological Relationship Diagram

In: 5th International Symposium on Innovative Nuclear Energy Systems (INES-5)

Tokyo, Japan

  • 31.10.-02.11.2016 (2016)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • Rui Li
  • W. Maschek
  • A. Rineiski

Bell-Plesset Instability Analysis for an Inward Centralized Sloshing

In: 17th International Conference on Emerging Nuclear Energy Systems (ICENES'2015)

Istanbul, Turkey

  • 04.-08.10.2015 (2015)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • Rui Li
  • W. Maschek
  • A. Rineiski

Bell-Plesset Instability Analysis for an Inward Centralized Sloshing

In: 17th International Conference on Emerging Nuclear Energy Systems (ICENES'2015)

Istanbul, Turkey

  • 04.-08.10.2015 (2015)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • X.-N. Chen
  • A. Rineiski
  • F. Gabrielli
  • L. Andriolo
  • Rui Li
  • W. Maschek

SIMMER-III Parametric Studies of Fuel-Steel Mixing and Radial Mesh Effects on Power Excursion in ESFR ULOF Transients

  • (2015)
This paper deals with SIMMER-III once-through simulations of the first power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered initially in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is – 2$, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped " homogenization " reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the first power excursion after the sodium boiling.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
Vortrag

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

Numerical investigation of SIMMER code for fuel-coolant interaction

In: 17th International Conference on Emerging Nuclear Energy Systems (ICENES'2015)

Istanbul, Turkey

  • 04.-08.10.2015 (2015)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
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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. 292C pg. 164-176.

  • (2015)

DOI: 10.1016/j.nucengdes.2015.06.011

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

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

Macroscopic Pin Bundle Model and its Blockage Simulations . Invited Talk

In: SEARCH-MAXSIMA International Workshop

Karlsruhe

  • 07.-10.10.2014 (2014)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • M. Massone
  • S. Wang
  • Rui Li
  • A. Rineiski

Safety Analysis of Liquid Fuel Fast Reactor System

In: 3rd ITU - INE Research Fellow Day

European Commission JRC Institute for Transuranium Elements Karlsruhe

  • 06.02.2014 (2014)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • E. Bubelis
  • G. Bandini
  • X.-N. Chen
  • I. Di Piazza
  • H. Doolaard
  • Rui Li

Fuel assembly blockage phenomena in a LFR: modeling approaches, assumptions, and results . Invited Talk

In: SEARCH-MAXSIMA International Workshop

Karlsruhe

  • 07.-10.10.2014 (2014)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • S. Wang
  • A. Rineiski
  • D. Zhang
  • E. Merle-Lucotte

Transient Analyses for Molten Salt Fast Reactor with Optimized Core Geometry

  • (2014)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

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

Fuel Dispersion after Pin Failure of MYRRHA-FASTEF Critical Core . Invited Talk

In: SEARCH-MAXSIMA International Workshop

Karlsruhe

  • 07.-10.10.2014 (2014)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

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

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

In: 4th International Symposium on Innovative Nuclear Energy Systems (INES-4)

Tokyo, Japan

  • 06.-08.11.2013 (2013)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

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

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

  • (2013)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

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

Subchannel Blockage Accident Analysis of MYRRHA Reactor

In: 4th Conference on Heavy Liquid Metal Coolants in Nuclear Technologies

Obninsk, Russian Federation

  • 22.-27.09.2013 (2013)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • M. Mori
  • H. Ninokata

Parametric Investigation on the Effect Factors for Liquid Droplet Impingement Erosion

  • (2011)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
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

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.
  • Europan Campus Rottal-Inn
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • H. Ninokata

Numerical Simulation of Impact Force Caused by Liquid Droplet Impingement in BWRs

  • (2010)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • H. Ninokata

Numerical Study on Turbulence Attenuation Model for Liquid Droplet Impingement Erosion in BWRs

  • (2010)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • E. Merzari
  • A. Yamaguchi
  • H. Ninokata
  • F. Watanabe
  • M. Mori

Computational Fluid Dynamics Study on Flow Modeling of Liquid Droplet Impingement Erosion in BWR

  • (2009)
The bent pipe wall thinning has been often found at the elbow of the drain lines in many BWRs. The wall thinning phenomena have been considered to be a subject of combined research areas of water chemistry and thermal hydraulics, primarily due to the corrosion and erosion mechanisms. Liquid droplet impingement erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the orifice flow of drain lines in BWRs. In this paper, three-dimensional numerical simulations are conducted for a bent pipe. Typically the pipe diameter is 170 mm and the bending angle is 90 degree; the mass flow rate of droplet is 4.53960x10-3 kg/s with the velocity of 280 m/s at the entry. The calculations employed 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-Stoke equations with RNG κ-ε turbulent governing equations 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 liquid 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. Finally, the effects of droplet size are also demonstrated numerically.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Beitrag in Sammelwerk/Tagungsband

  • Rui Li
  • E. Merzari
  • H. Ninokata

A numerical Study on Liquid Droplet Impingement Erosion in BWR

  • (2009)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

  • Y. Wang
  • Y.-L. He
  • Rui Li
  • Y.-G. Lei

Heat Transfer and Friction Characteristics for Turbulent Flow of Dimpled Tubes

In: Chemical Engineering & Technology vol. 32 pg. 956-963.

  • (2009)

DOI: 10.1002/ceat.200800660

The hydrodynamic and heat transfer characteristics of dimpled tubes with different arrangements were investigated in this paper. The dimples on the tube are outward-facing, raised, with aligned and staggered arrangements. The experiments were carried out in counter-current flow patterns with cold air along the tube inside and hot water around the outside of the tube. The Reynolds numbers of air flow were in the range of 1.6·104 to 5.4·104. By comparing the performances among three tubes (smooth tube, dimpled tube with aligned arrangement, and dimpled tube with staggered arrangement), we could find that the dimpled tubes enhanced the Nusselt number by about 18.6∼22.7 %, while the friction factors of the dimpled tubes increased by 18.6∼25.9 % for the aligned arrangement and by 14.3∼29.8 % for the staggered arrangement. The test results also showed that different dimple arrangements have very little effect, which is always within 5 %. By the integrated performance evaluation of (Nua/Nus)/(fa/fs)1/3, the maximum of the heat transfer enhancement with the same friction penalty, about 19 %, could be achieved by optimization of the dimpled-tube design.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

  • P. Chu
  • Y. He
  • Y. Lei
  • L. Tian
  • Rui Li

Three-dimensional numerical study on fin-and-oval-tube heat exchanger with longitudinal vortex generators

In: Applied Thermal Engineering vol. 29 pg. 859-876.

  • (2009)

DOI: 10.1016/j.applthermaleng.2008.04.021

Three-dimensional numerical study was performed for heat transfer characteristics and fluid flow structure of fin-and-oval-tube heat exchangers with longitudinal vortex generators (LVGs). For Re (based on the hydraulic diameter) ranges from 500 to 2500, it was found that the average Nu for the three-row fin-and-oval-tube heat exchanger with longitudinal vortex generators increased by 13.6–32.9% over the baseline case and the corresponding pressure loss increased by 29.2–40.6%. The results were analyzed on the basis of the field synergy principle to provide fundamental understanding of the relation between local flow structure and heat transfer augmentation. It was confirmed that the reduction of the intersection angle θ between the velocity field and the temperature field was one of the essential factors influencing heat transfer enhancement. Three geometrical parameters – placement of LVGs (upstream and downstream), angles of attack (α = 15°, 30°, 45° and 60°) and tube-row number (n = 2, 3, 4 and 5) – were also investigated for parameter optimization. The LVGs with placement of downstream, angles of attack α = 30° and minimum tube-row number provide the best heat transfer performance. The effects of the three geometrical parameters on heat transfer enhancement were also analyzed from the view point of the field synergy principle and it was found that the results can be well explained by the field synergy principle.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

  • Y.-L. He
  • Y.-G. Lei
  • W.-Q. Tao
  • J.-F. Zhang
  • P. Chu
  • Rui Li

Second-Law Based Thermodynamic Analysis of a Novel Heat Exchanger

In: Chemical Engineering & Technology vol. 32 pg. 86-92.

  • (2009)

DOI: 10.1002/ceat.200800396

In the present investigation, second-law based thermodynamics analysis was applied to a new heat exchanger with helical baffles. The helical baffles are designed as quadrant ellipses and each baffle occupies one quadrant of the cross-section of the shell side. Experimental tests were carried out with cold water in the tube side with a constant flow rate, and hot oil on the shell side with flow rate range from 4–24 m3/h. The temperatures and pressures for the inlet and outlet of both sides were measured. The heat transfer, pressure drop, entropy generation, and exergy loss of the new heat exchanger were investigated and compared with the results for a conventional shell-and-tube heat exchanger with segmental baffles. The computed results indicated that both the entropy generation number and exergy losses of the new heat exchanger design are lower than those of the heat exchanger with segmental baffles, which means that the novel heat exchanger has a higher efficiency than the heat exchanger with segmental baffles, from the second-law based thermodynamics viewpoint.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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)
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Vortrag

  • Rui Li
  • E. Merzari
  • H. Ninokata

A Numerical Study on Liquid Droplet Impingement Erosion in BWR

In: American Nuclear Society: 2009 Winter Meeting

Washington, D.C., USA

  • 15.-19.11.2009 (2009)
  • Europan Campus Rottal-Inn
  • NACHHALTIG
Zeitschriftenartikel

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

Design and optimization of heat exchangers with helical baffles

In: Chemical Engineering Science vol. 63 pg. 4386-4395.

  • (2008)

DOI: 10.1016/j.ces.2008.05.044

The hydrodynamics and heat transfer characteristics of a heat exchanger with single-helical baffles are studied experimentally as well as numerically. A heat exchanger with two-layer helical baffles is designed by using computational fluid dynamics (CFD) method. The comparisons of the performance of three heat exchangers with single-segment baffles, single-helical baffles and two-layer helical baffles, respectively, are presented in the paper. The experiment is carried out in counter-current flow pattern with hot oil in shell side and cold water in tube side. Overall heat transfer coefficients are calculated and heat transfer coefficients of shell side are determined by Wilson plots technique. It shows that the heat exchangers with helical baffles have higher heat transfer coefficient to the same pressure drop than that of the heat exchanger with segmental baffles based on the present numerical results, and the configuration of the two-layer helical baffles has better integrated performance than that of the single-helical baffles.
  • Europan Campus Rottal-Inn
  • NACHHALTIG
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

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.
  • Angewandte Naturwissenschaften und Wirtschaftsingenieurwesen
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Vita

  • 2018 – heute: Professor “Energietechnik – Gas-und Wärmenetze” an der Technischen Hochschule Deggendorf
  • 2012 – 2018: Wissenschaftlicher Mitarbeiter am Institut für Thermische Energietechnik und Sicherheit (bisher Institut für Kern- und Energietechnik) der Fakultät für Maschinenbau, Karlsruher Institut für Technologie (KIT)
  • 2008 – 2012: Doktorand im Laboratorium für fortgeschrittene Kernenergie der Fakultät für Maschinenbau, Tokyo Institute of Technology