Mohen Ghafari
Short Bio
I am an assistant professor with three years of experience in Department of Energy Engineering at Sharif University of Technology. I studied at Sharif University of Technology in all three levels and part of my Ph.D course at the University of Pisa. My field of research and work is focused on the processes and phenomena that occur in two-phase at power plant. Also a part of my activities is to develop industrial automation and the use of pneumatic systems.
Education
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Ph.D. 2019
Ph.D. in Department of Energy Engineering
Sharif University of Technology
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M.S. 2013
Department of Energy Engineering
Sharif University of Technology
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B.S. 2011
Department of Mechanical Engineering
Sharif University of Technology
Honors, Awards and Grants
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2016-2017Guest researcher in San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa,Pisa, Italy
Research Summary
My field of research is mainly focused on the power plants safety. In this field, a wide range of thermohydraulic phenomena emerges, which requires the use of system codes to analyze. As a result, improving the models used in computational fluid dynamics codes is one of my research activities. In addition, the analysis of operating events in the power plant is also one of my other research areas by examining the systems and sequence of events.
Interests
- Power Plant Technology
- Two-Phase Flow
- Heat Transfer
- Turbulence Modeling in CFD code
- Coupling CFD and system codes
Publications
This is a list of my publications so far.
Stress analysis of PWR pressurizer surge-line during single phase and two-phase thermal stratification
Journal
Paper Annals of Nuclear Energy,Marh 2023, Pages 55-67
Abstract
Non-uniform temperature distribution at horizontal pipes cross section leads to an asymmetric thermal stress. This condition would be considered as an ageing mechanism due to fatigue effect, known as thermal stratification. To evaluate this ageing mechanism, the stress analysis should be done in horizontal pipelines like surge-line connected the pressurizer to the hot leg of primary loop of NPPs by consideration of operational or transient conditions. During these conditions, the status of surge-line would be single phase or two-phase stratified flow. To estimate the equivalent stress in horizontal section of surge-line, the developed CFD model for both single and two-phase flow is coupled with mechanical structural analysis. In two phase-condition, the heat transfer coefficient is evaluated by employment of turbulent kinetic energy source function. The results show that the maximum stress is almost same for single and two-phase stratification, but the propagation of cold water layer intensifies the asymmetric stress in two-phase condition.
Axial-offset analysis in iPWR by developing the neutronic/thermal-hydraulic core simulator based on coarse-mesh methods
Journal
Paper Annals of Nuclear Energy,Feb 2023, Pages 55-67
Abstract
Axial-Offset (AO), as an important operational parameter, should be monitored in the load-following procedure due to the safety limits related to xenon oscillation and axial peaking factor in a nuclear power plant. In this study, to calculate AO for a Small Modular Reactor, a neutronic/thermal–hydraulic simulator has been developed based on coarse mesh methods. In this simulator, the average current nodal expansion method and the single heated channel approach have been used for neutronic and thermal–hydraulic analysis, respectively. The developed simulator is verified by the IAEA-3D model and thermal–hydraulic benchmark model. As the iPWR small modular reactor case study, SMART reactor has been simulated, and the neutronic and thermal–hydraulic parameters have been calculated. For investigation of AO behavior during power changes, the regulating bank repositioning approach is employed, and the reactor core characteristics are estimated by a developed simulator. The results demonstrate that the AO variation remains in a safe and standard operating region for SMART reactor.
Detection and estimation of faulty sensors in NPPs based on thermal-hydraulic simulation and feed-forward neural network
Journal paper Annals of Nuclear Energy, 2022, Volume
166
Abstract
Sensors are one of the most vital instruments in Nuclear Power Plants (NPPs), and operators and safety systems monitor and analyze various parameters reported by them. Failure to detect sensors malfunctions or anomalies would lead to the considerable consequences. In this research, a new method based on thermal–hydraulic simulation by RELAP5 code and Feed-Forward Neural Networks (FFNN) is introduced to detect faulty sensors and estimate their correct value. For design an efficient neural net, seven feature selectors (i.e., Information gain, ReliefF, F-regression, mRMR, Plus-L Minus-R, GA, and PSO), three sigmoid activation functions (i.e., Logistic, Tanh and Elliot), and three training algorithms (i.e., Levenberg–Marquardt (LM), Bayesian Regularization (BR) and Scaled Conjugate Gradient (SCG)) have been comprehensively compared and evaluated. The required data have been obtained by simulating LOFA and SBLOCA transients in RELAP5 code for the Bushehr Nuclear Power Plant (BNPP). The main advantage of this method is that with the failure of more than one sensor, the detection of other sensors is not completely disrupted, and are monitored continually and independently.
Prediction of steam/water stratified flow characteristics in NPPs transients using SVM learning algorithm with combination of thermal-hydraulic model and new data mapping technique
Journal Paper Annals of Nuclear Energy, Feb
2022, Volume 166
Abstract
Steam/water stratified flow would occur in transient condition (e.g. LOCA) in light water Nuclear Power Plants (NPPs). Due to high gradient of flow characteristics at the interface of steam/water flow, the prediction of flow characteristics (e.g. temperature, pressure, velocity, and Turbulent Kinetic Energy (TKE)) requires further attention and special interfacial models. Also, accurate simulation of these mentioned characteristics needs fine spatial mesh and very small time steps based on Computational Fluid Dynamics (CFD) standard criteria. In order to reduce the computational cost, the combination of thermal–hydraulic modelling and soft computing is considered as a new strategy in this study. The steam/water stratified flow in a rectangular channel (case 3 of Lim et al test section) is examined as case study and calculated values of the characteristics by thermal–hydraulic model are fed as training/test data to the Support Vector Machine (SVM) learning algorithm. SVM in combination with the proposed data mapping technique which is a type of autocorrelation finding predicts the value of each characteristic at a specific position/ time using the value of that characteristic at previous time at that position and previous position. The results show that the proposed methodology is appropriate for prediction of steam/water flow characteristics. Velocity, temperature, and TKE are predicted with reasonable accuracy. The predicted pressure shows a trend similar to the values obtained from the thermal–hydraulic modelling. For precise prediction of parameters similar to the pressure, it seems deep learning in combination with the proposed data mapping technique and a kind of features selection technique are needed. This method is under development and will be reported as the subsequent.
New turbulence modeling for air/water stratified flow
Journal
Paper Journal of ocean Engineering and Science,Marh 2020, Pages 55-67
Abstract
The prediction of interfacial turbulence characteristics is one of the still challenging of two-phase stratified flow. The evaluation of some important parameters such as interfacial heat transfer coefficient based on turbulence kinetic energy and turbulence dissipation rate in some models, intensifies the importance of turbulence flow correct simulation. High gradient of velocity and turbulence kinetic energy at the interface of two-phase stratified flow leads to a major overestimation or underestimation of flow characteristics without any special treatment. Consideration of a source function of turbulence eddy frequency at the interface is one of the common solution employed in past researches. Although this solution remedies some shortcomings of traditional methods in smooth stratified flow, its application in wavy stratified flow needs the other modifications. The examination of turbulence characteristics near the free surface reveals that, in addition to turbulence eddy frequency, the other source function of turbulence kinetic energy should be considered near the free interface. So, a new source function of turbulence kinetic energy is proposed at the interface based on flow condition. This new method has been employed for Fabre et al. (1987) experiment designed for air/water stratified flow. The results of simulation have a good agreement with experimental data and turbulence characteristic can be captured near the free surface.
New turbulence modeling for simulation of Direct Contact Condensation in two-phase pressurized thermal shock
Journal paperProgress in Nuclear Energy, Sep 2018, Page 358-371
Abstract
Injection of Emergency Core Cooling System (ECCS) water into the primary loops of the Pressurized Water Reactors (PWRs) leads to rapid cooling of Reactor Pressure Vessel (RPV) inside wall after Loss Of Coolant Accident (LOCA). This condition, known as Pressurized Thermal Shock (PTS) intensifies the propagation of the RPV structural defects and would be considered as an ageing mechanism. For structural and fracture analysis of RPV wall, thermal-hydraulic analysis of PTS should be accomplished to obtain the steam/water flow characteristics in the downcomer. For this purpose, simulation of steam/water stratified flow (due to density difference) after the injection point should be done by Computational Fluid Dynamics (CFD) methods. In this region, steam condensation over water layer is considered as the only heat source and controlled by turbulence eddy motion near the steam/water interface. Based on Surface Renewal Theory (SRT), Heat Transfer Coefficient (HTC) would be calculated by evaluation of turbulence length and velocity. Therefore, prediction of turbulence characteristics plays a significant role for estimation of interfacial mass transfer and temperature profile. High gradient of velocity and Turbulence Kinetic Energy (TKE), and interfacial mass and momentum transfer at the steam/water interface needs some modifications for application of traditional turbulence models. Implementation of damping function is one of the common solutions to overcome the overestimation of TKE at the steam/water interface. Although, this function improves flow characteristics of smooth stratified flow, investigation of conservation equations and experimental data implies that the other source function is needed when the flow regime changes to wavy flow. In this paper, a new source function of TKE based on variations of turbulence characteristics is proposed for steam/water interface leading to a special boundary condition of turbulence. To investigate the effects of this modification, simulation of air/water and steam/water stratified flow in three different test facilities is performed. The results show that the implementation of the source function of TKE improves the prediction of turbulence characteristics at the interface of isothermal stratified flow. Also condensation rate and temperature gradient of steam/water stratified flow have a better agreement with experimental data.
Boundary identification between LBLOCA and SBLOCA based on stratification and temperature gradient in two-phase PTS
Journal
Paper Annals of Nuclear Energy, May
2018, Pages 430-441
Abstract
Temperature gradient on the thick Reactor Pressure Vessel (RPV), caused by sudden overcooling events, especially in the downcomer, would intensify the propagation of structural defects. This situation known as Pressurized Thermal Shock (PTS) could be created in case of Emergency Core Cooling System (ECCS) actuation which leads to injection of cold water into the cold leg of the primary loop in some accidents, e.g. Loss Of Coolant Accident (LOCA). Prediction of Plant response to LOCA and water temperature gradient in the downcomer are performed in thermal-hydraulic section of PTS analysis. Employment of system codes is one of the proposed procedures in literature to obtain plant response and flow condition in the cold leg during LOCA. Also the results of these codes would be used to find the flow regime in the cold leg with some limitations. In this paper simulation of different break sizes in Bushehr Nuclear Power Plant as VVER-1000 reactor is performed by RELAP system code to find the temperature gradient and flow regime in the cold leg according to different criteria. Due to some limitations of system codes, CFX code is employed to evaluate turbulence characteristics at the interface for identification of flow regime. The comparison between results of different LOCA scenarios reveals a sharp reduction of water temperature in downcomer for large breaks which would be used for classification of LOCA. Also the flow regime in the cold leg during ECCS injection changes from stable stratified flow to wavy flow when the break size increases beyond a certain value. Therefore, the difference of temperature gradient in downcomer and flow regime in cold leg will be proposed as a new definition of Small Break LOCA (SBLOCA) and Large Break LOCA (LBLOCA) relevant to PTS analysis.
Professional Experience
Experiences Brief
I am adept at using TH system and CFD codes.I am currently more involved with scientific computations and statistical analysis.
Skills
- ANSYS (CFX,Fluent,Mechanical,TurboGrid,Vista<
- Matlab
- CFDTurbo
- Relap system code
- Fatek & Delta PLC
Contact & Meet Me
I would be happy to talk to you if you need my assistance in your research or anything else. I prefer to communicate through my Sharif email, but feel free to contact me however you like.
- 021-66166110
- Ghaffari@sharif.edu
- Ghaffari@sharif.ir
At My Office
You can find me at my office located at Second floor of Department of Energy Engineering .I am at my office on Sat and Mon, but you may consider an email to fix an appointment.