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In this work, the problem of re-ejection of a permeating droplet through a membrane pore back to the feed channel when the transmembrane pressure (TMP) becomes zero is investigated. This problem is important in the context of oily water filtration using membranes. In particular, in the novel periodic feed pressure technique (PFPT), which has been proposed to combat membrane fouling, the TMP alternates between the operating value and zero in a periodic manner. During the period in which TMP is high, filtration occurs, and when it is zero, cleaning commences. We are particularly interested in what happens to a droplet, initially undergoing permeation, when the TMP becomes zero. It is evident that when the TMP is zero the meniscus inside the pore reverses its motion toward the feed channel rather than toward the permeate side by the action of interfacial tension force. A theoretical model is built to determine the rate at which the meniscus inside the pore advances when the TMP is zero. The conservation of momentum equation is used to establish a one-dimensional model that updates the location of the meniscus with time. The derived model considers both quasi-static and dynamic scenarios. In addition, the model accounts for both the viscosity contrast between the two fluids, as well as the gravity. A computational fluid dynamics (CFD) simulation has been built to provide a framework for model verification and validation. The model, based on quasi-static conditions, provides an overall similar trend to that obtained via CFD analysis. Nevertheless, the quasi-static model predicts a more rapid meniscus advancement inside the pore than the CFD simulation. When the dynamic contact angle is incorporated, very good matching is observed.

Hydraulic transient is a flow situation where the velocity of the flow and the pressure change rapidly with time in pipes networks filled with fluid. Water hammer is one of the consequences that may occur accidentally in the reactors cooling system. The pressure level can exceed the limits given by the manufacturer, and it leads to the failure of the pipes integrity or nuclear fuel elements fixation in the core. Cooling system involves many components, and when a flow control component changes abruptly its status (for example, a valve closing or pump trip), it causes rapid changes in velocity and propagates pressure pulses in short time. The cooling system is equipped with flapper valve and flywheel which overcomes the transient issues. A pump or check valve may be stopped or closed abruptly and may cause bad impacts on the system which is simulated with soft trip using flywheel and without flywheel in case of its failure for any accidental reasons. In addition the abrupt valve closure is deemed as a major issue in the system if occurred. AFT Impulse 5.0 software is used to simulate and study the transient of each case. The simulation appears that the reactor cooling system may suffer from surge that may occur during the sudden closure of check valve but in other hand it can be avoided by regular maintenance of flapper valve to open in timely manner.

Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions. Fuel performance is a complicated phenomenon that involves thermal, mechanical, and irradiation mechanisms and requires special multiphysics modules. In this study, a fuel performance model was developed using the COMSOL Multiphysics platform. The modeling was performed for a 2D axis-symmetric geometry of a UO2 fuel pellet in the E110 clad for VVER-1200 fuel. The modeling considers all relevant phenomena, including heat generation and conduction, gap heat transfer, elastic strain, mechanical contact, thermal expansion, grain growth, densification, fission gas generation and release, fission product swelling, gap/plenum pressure, and cladding thermal and irradiation creep. The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power, in addition to validation of fission gas release (FGR) predictions. This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters. A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters. The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup, including the temperature profiles, gap width, fission gas release, and plenum pressure. A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models. Although the model was developed for normal operating conditions, it can be modified to include off-normal operating conditions.

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MWth. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MWth.

In this study, Copper-Polyaniline nanocomposite (Cu-PAni) has been synthesized by gamma radiolysis as a novel method and was confirmed by different analytical techniques such as Scanning Electron Microscope (SEM), Transmission Electron Microscope (TEM), X-ray powder diffraction (XRD). The copper nanoparticle completely covered by a layer of polyaniline with different thickness. The particle mean size was determined from XRD by using the Debye-Scherrer’s formula and the particle size was found to be 21 nm. The thermal conductivity of the nanofluid has been measured at different nanofluids volume fractions of 0.2%, 0.4%, 0.6%, 0.8% and 1.0%. The effect of nanofluid temperature on the nanofluid properties is measured in the range from 10 °C to 90 °C. The thermal conductivity ratio of the Copper-Polyaniline nanocomposites water nanofluid increases 107% at nanofluid temperature of 10 °C and 159% at 90 °C for a volume fraction of 1.0%. The Cu-PAni nanocomposite is a potential material for thermal conductivity enhancement of water based nanofluid.

Heat transfer study of nanofluids as coolant in SCWRs core has been performed at Helwan University. A thermal hydraulic code has been produced to study the effect of TiO2 nanofluid water based as a...

Heat transfer deterioration (HTD) phenomena in supercritical water is one of the most thermal hydraulic critical issues for supercritical reactors. The onset of heat transfer deterioration and bouncy effect have been studied in details, however limited work has been done on the mechanism of HTD occurrence. In this work, three dimensional analysis of velocity and temperature boundary layers under heat transfer deterioration conditions is conducted. The effect of thermo physical water properties changes inside the fluid boundary layer and its effect on the heat transfer coefficient is studied. Analogy between HTD and critical heat flux condition in subcritical boiling is observed and reported. Under HTD conditions, clotting or sudden stopping of the flow inside the tube is predicted due to the severe variation in density and other thermo physical properties in the tubes radial direction.

One important phenomenon occurring during Loss of Coolant Accident (LOCA) is Counter-Current Flow Limitation (CCFL). The incidence of such CCFL is introduced by the onset of hydraulic jump. In the present work, a one dimensional model was modified to fit circular hot channel. The model was used to study the factors affecting the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical coolant flow depth during circular horizontal hot leg injection in US-APWR Mitsubishi Reactor. The results showed good agreement with published experimental data of the Upper Plenum Test Facility (UPTF) at Mannheim, Germany. It was found that higher injected coolant mass flow rate increases the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical injection depth divided by the diameter of the injection pipe. Such behavior is thought to be due to the increase of the inertia force by increasing of the injected coolant mass flow rate and the inverse of the diameter of the injection pipe. It was found also that, the location of the occurrence of hydraulic jump increases with decreasing load effect. Therefore, these results reveal that the avoidance of CCFL as well as hydraulic jump through hot leg at maximum load can be achieved by decreasing the distance between the injection point and the pressure vessel to below 0.3 m, and with diameter of 4 in (10.16 cm) as the design diameter of the injection pipe in US-APWR Mitsubishi Reactor. Moreover, the maximum critical depth (56 cm) is less than the diameter of the hot leg (78.74 cm) at an injected coolant mass flow of 400 kg/s, and with diameter of 4 in (10.16 cm) as the design diameter of the injection pipe in US-APWR Mitsubishi Reactor; such result assure that there is no flow blockage in the hot leg yet.

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 80 cm length, 7 cm width and 2.7 mm gap thickness under different heat fluxes ranging from 2.7 kW/m2 to 26.8 kW/m2 that covers almost all possible heat fluxes in the single-phase liquid. The measured local Nusselt number values are compared with the predictions of two correlations for both natural and combined natural and forced convection regimes respectively. The comparison shows the need for developing new correlation to represent the obtained experimental results and to be utilized for the prediction of the local Nusselt number in thermal–hydraulics and safety analysis of the reactor. The developed correlation fits the experimental data within 5.8% relative standard deviation.