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Suche nach „[A.] [Rineiski]“ hat 10 Publikationen gefunden
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    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    Nuclear Engineering and Design, vol. 341, no. 146-154

    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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    Annals of Nuclear Energy, vol. 110, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    Zeitschriftenartikel

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

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

    Annals of Nuclear Energy, vol. 104, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    Zeitschriftenartikel

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

    Numerical investigation of SIMMER code for fuel-coolant interaction

    International Journal of Hydrogen Energy, vol. 41, no. 17, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    International Journal of Hydrogen Energy, vol. 41, no. 17, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    Zeitschriftenartikel

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

    Bell-Plesset Instability Analysis for an Inward Centralized Sloshing

    Nuclear Engineering and Design, vol. 297, pp. 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%.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    Zeitschriftenartikel

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

    Macroscopic pin bundle model and its blockage simulations

    Energy Conversion and Management, vol. 91, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    Annals of Nuclear Energy, vol. 79, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    Energy Procedia, vol. 71, pp. 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.

    NachhaltigAngewandte Naturwissenschaften und WirtschaftsingenieurwesenEuropan Campus Rottal-Inn

    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

    Nuclear Engineering and Design, vol. 292, no. October, pp. 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.