Salah El-Din El-Morshedy

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Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool

Hany Khater, Salah El-Din EL-Morshedy, Abdelfatah Abdelmaksoud

Annals of Nuclear Energy > 2015 > 83 > C > 376-385

CFD is expected to feature more frequently in reactor thermal hydraulics. The reason for the increased use of multidimensional CFD methods is not only the increased availability of capable computer systems but also the ongoing drive to improve and reduce uncertainty in our predictions of important phenomena. In this work, a CFD model coupled with the reactor point kinetics equations is developed using...

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Modeling and simulation of loss of the ultimate heat sink in a typical material testing reactor

Hisham El-Khatib, Salah El-Din El-Morshedy, Maher.G. Higazy, Karam El-Shazly

Annals of Nuclear Energy > 2013 > 51 > Complete > 156-166

A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in a typical material testing reactor (MTR). The model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The model is validated against PARET code for steady-state operation and verified by the reactor operation records for transients. Then, the model is used to...

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Thermal-hydraulic modeling and analysis of a tank in pool reactor for normal operation and loss of flow transient

Salah El-Din El-Morshedy

Progress in Nuclear Energy > 2012 > 61 > Complete > 78-87

A transient thermal-hydraulic model entitled Tank in Pool Reactor Thermal-Hydraulic Analysis (TPRTHA) has been developed to simulate the steady-state operation and loss of flow transient for a tank in pool type research reactor. The model solves the momentum equation, energy equation and general conduction equation in cylindrical coordinates in order to predict the coolant velocity, coolant temperature...

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CFD analysis of flow blockage in MTR coolant channel under loss-of-flow transient: Hot channel scenario

Amgad Salama, Salah El-Din El-Morshedy

Progress in Nuclear Energy > 2012 > 55 > Complete > 78-92

In this work 2D, CFD analysis of the fast loss-of-flow accident in the IAEA 10MW generic research reactor is conducted under the hot channel condition with one channel subject to a sequence of area contractions up to 80%. While the previous work on this system under normal operating conditions shows that the system may be safe up to blockage ratio of approximately 85%, this work shows that boiling...

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CFD simulation of flow blockage through a coolant channel of a typical material testing reactor core

Amgad Salama, Salah El-Din El-Morshedy

Annals of Nuclear Energy > 2012 > 41 > Complete > 26-39

Two-dimensional simulations of partial and total channel blockage of the IAEA 10MW generic reactor is performed under both hot and average channel conditions using the CFD code, Fluent. Three adjacent coolant channels of a fuel assembly and their fuel plates are modeled, one of which is subjected to partial blockage as a sequence of continuous area contractions due to buckling of its fuel-plates inwards...

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Predictive study of the onset of flow instability in narrow vertical rectangular channels under low pressure subcooled boiling

Salah El-Din El-Morshedy

Nuclear Engineering and Design > 2012 > 244 > Complete > 34-42

One of the most limiting critical phenomena, which nuclear designers must consider, is the onset flow instability (OFI) phenomenon. It is the critical phenomenon limiting the reactor power from the thermal-hydraulic point of view especially material testing reactors (MTR). In the present study, an empirical correlation is developed to predict OFI in narrow vertical rectangular channels simulating...

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Experimental investigation of natural convection heat transfer in narrow vertical rectangular channel heated from both sides

Salah El-Din El-Morshedy, Adel Alyan, Loula Shouman

Experimental Thermal and Fluid Science > 2012 > 36 > Complete > 72-77

The heat transfer characteristics of the natural convection regime through a vertical rectangular channel simulating a cooling channel of typical material testing reactor have been experimentally investigated. Experiments are performed on demineralized water as coolant passing under atmospheric pressure through narrow rectangular channel of 80cm length, 7cm width and 2.7mm gap thickness under different...

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CFD simulation of the IAEA 10MW generic MTR reactor under loss of flow transient

Amgad Salama, Salah El-Din El-Morshedy

Annals of Nuclear Energy > 2011 > 38 > 2-3 > 564-577

Three-dimensional simulation of the IAEA 10MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow...

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Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling

Salah El-Din El-Morshedy

Nuclear Engineering and Design > 2011 > 241 > 1 > 226-235

Research reactors of power greater than 20MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic...

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Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

Salah El-Din El-Morshedy, Ahmed Hassanein

Fusion Engineering and Design > 2010 > 85 > 5 > 687-693

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by...

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Transient thermal hydraulic modeling and analysis of ITER divertor plate system

Salah El-Din El-Morshedy, Ahmed Hassanein

Fusion Engineering and Design > 2009 > 84 > 12 > 2158-2166

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both...

article

Thermal–hydraulic modeling of the onset of flow instability in MTR reactors

Hany A. Khater, Salah El-Din El-Morshedy, Mohamed M.A. Ibrahim

Annals of Nuclear Energy > 2007 > 34 > 3 > 194-200

Prediction of the onset of the flow instability (OFI) in steady and transient sub-cooled flow boiling is an important consideration in the design and operation of nuclear reactors, in particular for materials testing reactors (MTR). In this study, a predictive model for OFI in the MTR has been developed. The model is based on both the heat balance during the bubble generation and condensation processes,...

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Search results for: Salah El-Din El-Morshedy

Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool

Modeling and simulation of loss of the ultimate heat sink in a typical material testing reactor

Thermal-hydraulic modeling and analysis of a tank in pool reactor for normal operation and loss of flow transient

CFD analysis of flow blockage in MTR coolant channel under loss-of-flow transient: Hot channel scenario

CFD simulation of flow blockage through a coolant channel of a typical material testing reactor core

Predictive study of the onset of flow instability in narrow vertical rectangular channels under low pressure subcooled boiling

Experimental investigation of natural convection heat transfer in narrow vertical rectangular channel heated from both sides

CFD simulation of the IAEA 10MW generic MTR reactor under loss of flow transient

Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

Thermal–hydraulic modeling of the onset of flow instability in MTR reactors

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