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Albuquerque, NM, United States

El-Genk M.S.,Institute for Space and Nuclear Power Studies
International Journal of Nuclear Energy Science and Technology | Year: 2010

Nuclear power is a mature industry with an incredible record of safety and reliability, without the emission of greenhouse gases, and one that is becoming economically attractive for private investment. The current interest in nuclear power for meeting future electricity and seawater desalination needs in the Gulf Corporation Council (GCC) states and other countries in the Middle East is prudent, logical and timely. An achievable goal for the GCC states would be to secure 30% of their future needs of electricity and of process heat for industrial applications and seawater desalination from nuclear power by 2030. This is equivalent to completing the construction of two 1500 MWe nuclear power plants each year starting in 2016. Accomplishing this goal requires a multifaceted approach to addressing many challenges that include: • stimulating private investments in nuclear energy • establishing specialised higher education and vocational training programmes • maintaining close cooperation with the International Atomic Energy Agency (IAEA) and countries with advanced nuclear capabilities • establishing a viable heavy industry • collaborating with other countries in the Middle East on the development of a common electric grid • investing in the mining and exploration of uranium resources • identifying and licensing suitable sites for construction of nuclear plants • establishing a regulatory and safety board that has a government oversight and an effective technical and R&D infrastructure • considering standardisation versus diversification in nuclear reactor types. Copyright © 2010 Inderscience Enterprises Ltd. Source

El-Genk M.S.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,University of New Mexico
Heat Transfer Engineering | Year: 2012

Enhancements in nucleate boiling heat removal with dielectric liquids, by increasing either the bubbles nucleation sites density and/or the wetted surface area, are desirable for immersion cooling of high-power computer chips. This article presents the results of recent investigations of nucleate boiling enhancement of FC-72, HFE-7100, and PF-5060 dielectric liquids on porous graphite, copper microporous surfaces, and copper surfaces with square corner pins, 3 mm × 3 mm in cross-section and 2, 3, and 5 mm tall. All surfaces have a footprint measuring 10 × 10 mm. These investigations examined the effects of liquid subcooling up to 30 K and surface inclination, from upward-facing to downward-facing, on nucleate boiling heat transfer coefficient and critical heat flux. Natural convection of dielectric liquids for cooling chips while in the stand-by mode, at a surface average heat flux <20 kW/m 2, is also investigated for the different surfaces. © 2012 Copyright Taylor and Francis Group, LLC. Source

El-Genk M.S.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,University of New Mexico | Tournier J.-M.P.,Institute for Space and Nuclear Power Studies
International Congress on Advances in Nuclear Power Plants 2010, ICAPP 2010 | Year: 2010

A massive air or steam ingress in High and Very-High Temperature Reactors (HTRs and VHTRs) nominally operating at 600-950 °C is a design-basis accident requiring the development and validation of models for predicting the graphite oxidation and erosion and examining the potential of a fission products release and a loss in integrity of the graphite core and reflector blocks. Isotropic and porous nuclear graphite is of many types with similarities but also differences in microstructure; volume porosity, impurities; type and size of filler coke particles; graphitization; heat treatment temperature and thermal and physical properties. These as well as temperature affect the prevailing mode and kinetics of the graphite oxidation and burn-off rate. This paper reviews the fabrication procedures, characteristics, chemical kinetics and modes of oxidation of nuclear graphite for future model developments. Source

Travis B.W.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,University of New Mexico
Fusion Science and Technology | Year: 2012

This paper presents the results of a coupled 3-D thermalhydraulics and CFD analysis of helium flow in a coolant channel of a prismatic core, Very High Temperature Reactor (VHTR). Results are used to develop a turbulent convection heat transfer correlation that accounts for induced mixing in the entrance region as: h=[0.10(k/D)Re b 0.653Pr b .0.4][1 + 0.57 e -(0.20z/D)] The entrance effect (second term) increases the local turbulent heat transfer coefficient, but diminishes for z/D > 25. This correlation is within ± 2% of the numerical results. Source

Palomino L.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,University of New Mexico
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015 | Year: 2015

This paper presents the results of 3-D Computational Fluid Dynamic (CFD) and thermal-hydraulic analyses investigating passive decay heat removal for the Scalable Liquid Metal cooled small Modular (SLIMM) reactor by natural circulation of ambient air, in case of a malfunction of the in-vessel helically coiled tubes, Na/Na heat exchanger. Results show that the longitudinal metal fins along the outer surface of the reactor guard vessel effectively increase the heat removal by air natural circulation. The thermal radiation from the guard vessel outer wall and metal fins is a major contributor to the heat removal of the decay heat by ambient air, accounting for 29%-42% of the total rate of heat removal. Results showed that the decay heat removal by ambient air is quite effective, even without metal fins along the outer surface of the guard vessel wall (∼ 1.0 MWth). The metal fins increase the rate of the heat removal by naturally convection of ambient air by an additional 26% to 1.26 MW4. Without metal fins along the outer surface of the guard vessel wall, the average temperature of the circulating liquid sodium in the reactor primary vessel peaks at ∼ 821.7 K,∼ 1.5 hr after reactor shutdown and decreases to 400 K ∼ 22.2 hr after reactor shutdown. With metal fins, the higher rate of heat removal (1.26 MWth), limits the peak temperature of the in-vessel liquid sodium to ∼ 806 K, only ∼ 40 minutes (or 0.665 hr) after reactor shutdown. These results confirm a large safety margin, > 330 K, from the boiling temperature of liquid sodium (∼ 1156 K at 0.1 MPa). Source

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