Albuquerque, NM, United States
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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.

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.

El-Genk M.S.,University of New Mexico | El-Genk M.S.,Institute for Space and Nuclear Power Studies | Tournier J.-M.P.,University of New Mexico | Tournier J.-M.P.,Institute for Space and Nuclear Power Studies | And 2 more authors.
Journal of Propulsion and Power | Year: 2010

A dynamic simulation model for space reactor power systems with multiple closed Brayton cycle loops for energy conversion is developed and demonstrated for a startup transient. The simulated power system employs a submersion-subcritical safe space S ^ 4 reactor with a negative temperature reactivity feedback and has no singlepoint failures in reactor cooling and energy conversion. The S ∧ 4 reactor core is divided into three hydraulically independent sectors, and each sector has a separate closed Brayton cycle loop that is thermal-hydraulically coupled to a circulating liquid NaK-78 secondary loop with two water heat pipe heat rejection radiator panels. Each closed Brayton cycle loop has a Brayton rotating unit designed and optimized for high thermal efficiency and low specific mass (116 kg/kWe). The reactor coolant and the closed Brayton cycle working fluid is a He-Xe binary gas mixture with a molecular weight of 40 g=mol. Results are presented for a startup transient of the S ^ 4 closed Brayton cycle power system to full-power operation at a reactor thermal power 471 kWth, a Brayton rotating unit shaft speed of 45 krpm, and turbine and compressor inlet temperatures of 1149 and 400 K, respectively. At these conditions, the nominal electrical power and thermal efficiency of the power system are 130.8 kWe and 27.8%. © 2010 by the American Institute of Aeronautics and Astronautics, Inc.

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.

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.

Rodriguez S.B.,Sandia National Laboratories | Rodriguez S.B.,Institute for Space and Nuclear Power Studies | El-Genk M.S.,Institute for Space and Nuclear Power Studies
Nuclear Engineering and Design | Year: 2010

The helium-cooled, high temperature Next Generation Nuclear Plant (NGNP) and Very High Temperature Reactor (VHTR) with prismatic type cores are being designed to operate at reactor exit temperatures ranging from 873 to 923 K and 1123 to 1223 K, respectively. The helium flow velocity in the coolant channels of the core is ∼70 m/s. The high-temperature helium jets exiting the coolant channels impinge onto the bottom plate in the lower plenum (LP), causing "hot spots" ("hot streaking") and stratification due to the absence of proper mixing and the obstruction caused by the graphite support columns. In order to minimize or eliminate hot streaking and enhance helium mixing in the LP, this work investigates using static, quadruple helicoid inserts at the exit of the coolant channels. These inserts introduce radial and azimuthal momentum flow components, which result in extensive entrainment and mixing of the surrounding gas in the LP, significantly reducing the impingement onto the bottom plate, thereby minimizing hot streaking and stratification. Results of parametric analyses and a comparison of the flow fields of helium free conventional and swirling jets, and of a convectional jet in cross flow are presented and discussed. The analyses with helium at 1273 K and the dynamic Smagorinsky turbulence model are conducted using Fuego, a 3D, finite element, incompressible, reactive flow, massively parallel code with state-of-the-art turbulence models developed at Sandia National Laboratories. The calculations are benchmarked successfully by comparing the numerical results with experimental data and semi-empirical analytical expressions. © 2010 Elsevier B.V. All rights reserved.

El-Genk M.S.,Institute for Space and Nuclear Power Studies | Ali A.F.,Institute for Space and Nuclear Power Studies
Journal of Heat Transfer | Year: 2015

Pool boiling experiments are performed to investigate potential enhancement of critical heat flux (CHF) of PF-5060 dielectric liquid on microporous copper (MPC) surfaces and the effect of surface inclination angle. The morphology and microstructure of the MPC surfaces change with thickness. The experiments tested seven 10 × 10 mm MPC surfaces with thicknesses from 80 to 230 μm at inclination angles of 0 deg (upward facing), 60 deg, 90 deg (vertical), 120 deg, 150 deg, 160 deg, 170 deg, and 180 deg (downward facing). CHF increases as the thickness of the surface increases and/or the inclination angle decreases. The values in the upward facing orientation are 36-59% higher than on smooth Cu. For all surfaces, CHF values in the downward facing orientation are approximately 28% of those in the upward facing orientation. A developed CHF correlation, similar to those of Zuber and Kutateladze, accounts for the effects of inclination angle and thickness of the MPC surfaces. It is in good agreement with experimental data to within ±8%. Still photographs of nucleate boiling on the MPC surfaces at different inclinations help the interpretation of the experimental results. © 2015 by ASME.

El-Genk M.S.,University of New Mexico | El-Genk M.S.,Institute for Space and Nuclear Power Studies | Gallo B.M.,University of New Mexico | Gallo B.M.,Institute for Space and Nuclear Power Studies
Journal of Propulsion and Power | Year: 2010

The University of New Mexico Brayton Rotating Unit-3 (UNM-BRU-3), designed for 40 g=mole He-Xe working fluid, is optimized for shaft speed of 45 kr pm, turbine and compressor inlet temperatures of 1149 and 400 K, and thermal power input of 157 kW th. At these conditions, the electrical power and thermal efficiency are 54:2 kW e and 34.5%, the compressor exit pressure is 1.044 MP a, the He-Xe flow rate is 1:54 kg=s, and the corresponding specific mass is 0:98 kg=kW e. Investigated are the effects of decreasing the turbine inlet temperature to 900 K and the thermal power input to 40 kW th and varying the shaft speed from 30 to 55 krpm. At 900 K turbine inlet temperature, thermal power input of 157 kW th, and shaft rotation speed of 45 krpm, the UNM-BRU-3 has a thermal efficiency of 22.8% and generates 34:2 kW e at a compressor exit pressure of 1.3 MP a and He-Xe flow rate of 2:07 kg/s;but the specific mass of the unit increases to ∼1:55 kg=kW e. The unit performance at 900 K is attractive for early deployment of space reactor and solar dynamics power systems with stainless steel structure, thus minimizing development cost and enhancing operation reliability and life. The low specific mass of UNM-BRU-3 will reduce the total mass and launch cost of these systems. © 2009 by M. S. El-Genk.

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).

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