Engineering topics

El-Genk, Mohamed S.

Publications and source records attributed to El-Genk, Mohamed S..

At least 19 records

DOE OSTI2023

Miniature DC electromagnetic pumps of molten lead and sodium to support development of Gen-IV nuclear reactors

This work developed designs of submerged, dual-regions, miniature DC-EM pumps for circulating molten lead and liquid sodium at ≤500 °C without active cooling for use in ex-pile and in-pile test loops to support of materials and fuel developments for Gen-IV sodium and molten lead fast nuclear reactor. These pumps with two pumping regions are 57 mm, 66.8 mm, 95.4 mm, and 133.5 mm in diameter and have two Alnico 5 permanent magnets with Hiperco-50 pole pieces for focusing the magnetic field lines in the flow duct. The Equivalent Circuit Model (ECM) linked to the Finite Element Method Magnetics (FEMM) software in MATLAB platform calculated the pumps characteristics for a wide range of parameters. Also determined are the pumps dimensions for the highest cumulative pumping power and peak efficiency. These dimensions are the height, width, and length of the flow duct, the thickness of the Alnico magnets, the length of current electrodes, and the separation distance between the two pumping regions. For molten lead, the pumping power increases from 368 W to 728 W, and the peak efficiency from 14.7% to 31.6% with increased pump diameter from 57 mm to 133.5 mm. In conclusion, for liquid sodium, both the pumping power and peak efficiency are higher, increasing from 392 W to 767 W and from 44.3 to 51.2%, respectively, with increased pump diameter from 57 mm to 133.5 mm.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Thermal analyses of high-power advanced thermoacoustic radioisotope power system for future space exploration missions

This paper presents the results of 3-D thermal analyses of the heat source assembly of a 440 We advanced radioisotope power system (RPS) for future space exploration missions. This high thermal efficiency RPS employs a heat source assembly of eight Step 2 General Purpose Heat Source (GPHS) modules and four Thermoacoustic Power Converter (TAPC) units. Each unit is thermally coupled to two GPHS modules. Here, the analyses investigated the temperatures of the Iridium (Ir) alloy cladding of the 238 PuO 2 fuel pellets and the surface of the Fine Weave Pierced Fabric (FWPF) aeroshell in the GPHS modules to ensure reliable operation and safety in case of an unlikely reentry. Conductive coupling of the GPHS modules to the heater heads of the TAPC units produces relatively uniform axial and azimuthal distributions of heat flux and temperature. However, for this case the temperature of the Ir alloy cladding is lower than the desired values of 1173 K to maintain sufficient ductility. Adding a 5 mm wide He filled gap and decreasing the thickness of the graphite sleeve to 10 mm increased Ir cladding temperature to 1197–1174 K, which are in the desired range. The maximum surface temperatures of the FWPF graphite aeroshell of 1022–1131 K is well below the NASA specified limit of 1373 K. In addition, the total mass of the heat source assembly decreased by 19.6 kg, ~43% saving compared to original design.

42 ENGINEERING↗

DOE OSTI2021

Pressurizer dynamic model and emulated programmable logic controllers for nuclear power plants cybersecurity investigations

This work demonstrates the functionality of the pressurizer using a fast-running three regions, non-equilibrium model and control programs of emulated PLCs. The state variables from an integrated model of primary loop in a representative PWR plant are communicated to the pressurizer’s emulated PLCs using a synchronized data transfer function. In turn, the PLCs communicate back instructions to the pressurizer model to adjust the pressure and water level to remain within preprogramed setpoints. Pressurizer model simultaneously solves the coupled mass and energy conservation equations in the saturated vapor and saturated and subcooled liquid regions using fixed step solver. Results demonstrate the response of the pressurizer model linked to an emulated pressure and water level PLCs in a simulated transient involving surge-in and surge-out events. The 50 ms response delay time for the emulated PLC insignificantly affects operation and predictions of the pressurizer model.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Extrapolation of thermal conductivity in non-equilibrium molecular dynamics simulations to bulk scale

Predictions of the bulk scale thermal conductivity of solids using non-equilibrium molecular dynamics (MD) simulations have relied on the linear extrapolation of the thermal resistivity versus the reciprocal of the system length in the simulations. Several studies have reported deviation of the extrapolation from linearity near the micro-scale, raising a concern of its applicability to large systems. To investigate this issue, present work conducted extensive MD simulations of silicon with two different potentials (EDIP and Tersoff-II) for unprecedented length scales up to 10.3 μm and simulation times up to 530 ns. For large systems ≥0.35 μm in size the non-linearity of the extrapolation of the reciprocal of the thermal conductivity is mostly due to ignoring the dependence of the thermal conductivity on temperature. To account for such dependence, the present analysis fixes the temperature range for determining the gradient for calculating the thermal conductivity values. However, short systems ≤0.23 μm in size show significant non-linearity in the calculated thermal conductivity values using a temperature window of 500 ± 10 K from the simulations results with the EDIP potential. Since these system sizes are shorter than the mean phonon free path in EDIP (~0.22 μm), the nonlinearity may be attributed to phonon transport. For the MD simulations with the Tersoff-II potential there is no significant non-linearity in the calculated thermal conductivity values for systems ranging in size from 0.05 to 5.4 μm.

42 ENGINEERING↗

NASA NTRS2005

Performance test results of a skutterudite-based unicouple with a metallic coating

A performance test of a Skutterudite-based unicouple (MAY-04) with a metallic coating to suppress the sublimation of antimony from the legs near the hot junction is performed in vacuum (~ 9 x10-7 torr) for ~ 2,000 hours at hot and cold junction temperatures of 892.1 + 11.9 K and 316.1 + 5.5 K, respectively. The p-leg is made of CeFe3.5Co0.5Sb12 and the n-leg is made of CoSb3. Presented are the measured voltage-current characteristics, electrical power, open-circuit voltage, and Seebeck coefficients of the legs as functions of cumulative test time. Also presented is the estimate of the conversion efficiency, shortly (~ 96 hrs) after the start of test. To demonstrate the effectiveness of the metallic coating, the measurements for MAY-04 are compared with those of two uncoated unicouples of the same leg materials (MAR-03 and JUN-03), which had been tested earlier. The cross-sectional areas of the legs in MAY-04 are larger than those in MAR-03 and JUN-03, tested in argon cover gas at ~ 0.051-0.068 MPa for 450 and 1200 hours, respectively. The open circuit voltage, Voc (204 mV) of MAY-04 at Beginning-Of-Test (BOT) is almost the same as that of MAR-03, but higher than that of JUN-03 (~180 mV). Although the argon gas effectively decreased antimony loss from legs of MAR-03 and JUN-03, marked degradations in performance occurred with time in these tests. Conversely, the metallic coating in MAY-04 effectively reduced the performance degradation with cumulative test time. The estimated peak efficiency of MAY-04, shortly after BOT (10.65%) is only ~ 0.37 percentage point lower than the theoretical value, assuming zero side heat losses and zero contact resistance per leg. The peak power of MAY- 04 decreased by only ~12%, from its BOT value of ~1.6 We to ~ 1.4 We after ~ 2,000 of cumulative testing

Caillat, Thierry↗

NASA NTRS1995

User's Manual for HPTAM: a Two-Dimensional Heat Pipe Transient Analysis Model, Including the Startup from a Frozen State

This report describes the user's manual for 'HPTAM,' a two-dimensional Heat Pipe Transient Analysis Model. HPTAM is described in detail in the UNM-ISNPS-3-1995 report which accompanies the present manual. The model offers a menu that lists a number of working fluids and wall and wick materials from which the user can choose. HPTAM is capable of simulating the startup of heat pipes from either a fully-thawed or frozen condition of the working fluid in the wick structure. The manual includes instructions for installing and running HPTAM on either a UNIX, MS-DOS or VMS operating system. Samples for input and output files are also provided to help the user with the code.

Tournier, Jean-Michel↗

NASA NTRS1995

HPTAM, a two-dimensional Heat Pipe Transient Analysis Model, including the startup from a frozen state

A two-dimensional Heat Pipe Transient Analysis Model, 'HPTAM,' was developed to simulate the transient operation of fully-thawed heat pipes and the startup of heat pipes from a frozen state. The model incorporates: (a) sublimation and resolidification of working fluid; (b) melting and freezing of the working fluid in the porous wick; (c) evaporation of thawed working fluid and condensation as a thin liquid film on a frozen substrate; (d) free-molecule, transition, and continuum vapor flow regimes, using the Dusty Gas Model; (e) liquid flow and heat transfer in the porous wick; and (f) thermal and hydrodynamic couplings of phases at their respective interfaces. HPTAM predicts the radius of curvature of the liquid meniscus at the liquid-vapor interface and the radial location of the working fluid level (liquid or solid) in the wick. It also includes the transverse momentum jump condition (capillary relationship of Pascal) at the liquid-vapor interface and geometrically relates the radius of curvature of the liquid meniscus to the volume fraction of vapor in the wick. The present model predicts the capillary limit and partial liquid recess (dryout) in the evaporator wick, and incorporates a liquid pooling submodel, which simulates accumulation of the excess liquid in the vapor core at the condenser end.

Tournier, Jean-Michel↗

NASA NTRS1994

A review of nuclear thermal propulsion carbide fuel corrosion and key issues

Corrosion (mass loss) of carbide nuclear fuels due to their exposure to hot hydrogen in nuclear thermal propulsion engine systems greatly impacts the performance, thrust-to-weight and life of such systems. This report provides an overview of key issues and processes associated with the corrosion of carbide materials. Additionally, past pertinent development reactor test observations, as well as related experimental work and analysis modeling efforts are reviewed. At the conclusion, recommendations are presented, which provide the foundation for future corrosion modeling and verification efforts.

Pelaccio, Dennis G.↗

NASA NTRS1993

Two-dimensional thermal-hydraulics analyses of the Pellet Bed Reactor for Nuclear Thermal Propulsion

Thermal-hydraulics design and analyses of the Pellet Bed Reactor for nuclear thermal propulsion are performed using the nuclear propulsion thermal-hydraulic analysis model to determine the 2D steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot-frit to avoid hot spots in the core at full power operation. Results show that by properly adjusting the axial porosity profile in the hot frit, hot spots in the core can be essentially eliminated during full power operation. This important accomplishment is achieved at the expense of slightly larger pressure losses in the core because of flow restriction at the hot frit. However, the overall pressure losses is only about 11 percent of the propellant inlet pressure.

Morley, Nicholas J.↗

NASA NTRS1992

Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

Juhasz, Albert J.↗

NASA NTRS1992

Natural circulation decay heat removal from an SP-100, 550 kWe power system for a lunar outpost

This research investigated the decay heat removal from the SP-100 reactor core of a 550-kWe power system for a lunar outpost by natural circulation of lithium coolant. A transient model that simulates the decay heat removal loop (DHRL) of the power system was developed and used to assess the system's decay heat removal capability. The effects of the surface area of the decay heat rejection radiator, the dimensions of the decay heat exchanger (DHE) flow duct, the elevation of the DHE, and the diameter of the rise and down pipes in the DHRL on the decay heat removal capability were examined. Also, to determine the applicability of test results at earth gravity to actual system performance on the lunar surface, the effect of the gravity constant (1 g and 1/6 g) on the thermal behavior of the system after shutdown was investigated.

El-Genk, Mohamed S.↗

NASA NTRS1992

Space nuclear power systems 1989; Proceedings of the 6th Symposium, Albuquerque, NM, Jan. 8-12, 1989. Vols. 1 & 2

The present conference discusses such space nuclear power (SNP) issues as current design trends for SDI applications, ultrahigh heat-flux systems with curved surface subcooled nucleate boiling, design and manufacturing alternatives for low cost production of SNPs, a lightweight radioisotope heater for the Galileo mission, compatible materials for uranium fluoride-based gas core SNPs, Johnson noise thermometry for SNPs, and uranium nitride/rhenium compatibility studies for the SP-100 SNP. Also discussed are system issues in antimatter energy conversion, the thermal design of a heat source for a Brayton cycle radioisotope power system, structural and thermal analyses of an isotope heat source, a novel plant protection strategy for transient reactors, and beryllium toxicity.

El-Genk, Mohamed S.↗

NASA NTRS1991

An investigation of voids formation mechanisms and their effects on freeze and thaw processes of lithium and lithium fluoride

The mechanisms of void formation during the cooldown and freezing of lithium coolant within the primary loop of SP-100 type systems are investigated. These mechanisms are: (1) homogeneous nucleation; (2) heterogeneous nucleation; (3) normal segregation of helium gas dissolved in liquid lithium; and (4) shrinkage of lithium during freezing. To evaluate the void formation potential due to segregation, a numerical scheme that couples the freezing and mass diffusion processes in both the solid and liquid regions is developed. The results indicated that the formation of He bubbles is unlikely by either homogeneous or heterogeneous nucleation during the cooldown process. However, homogeneous nucleation of He bubbles following the segregation of dissolved He in liquid lithium ahead of the solid-liquid interface is likely to occur. Results also show that total volume of He void is insignificant when compared to that of shrinkage voids. In viewing this, the subsequent research focuses on the effects of shrinkage void forming during freezing of lithium on subsequent thaw processes are investigated using a numerical scheme that is based on a single (solid/liquid) cell approach. The cases of lithium-fluoride are also investigated to show the effect of larger volume shrinkage upon freezing on the freeze and thaw processes. Results show that a void forming at the wall appreciably reduces the solid-liquid interface velocity, during both freeze and thaw, and causes a substantial rise in the wall temperature during thaw. However, in the case of Li, the maximum wall temperature was much lower than the melting temperature of PWC-11, which is used as the structure material in the SP-100 system. Hence, it is included that a formation of hot spots is unlikely during the startup or restart of the SP-100 system.

El-Genk, Mohamed S.↗

NASA NTRS1991

Conceptual studies on the integration of a nuclear reactor system to a manned rover for Mars missions

Multiyear civilian manned missions to explore the surface of Mars are thought by NASA to be possible early in the next century. Expeditions to Mars, as well as permanent bases, are envisioned to require enhanced piloted vehicles to conduct science and exploration activities. Piloted rovers, with 30 kWe user net power (for drilling, sampling and sample analysis, onboard computer and computer instrumentation, vehicle thermal management, and astronaut life support systems) in addition to mobility are being considered. The rover design, for this study, included a four car train type vehicle complete with a hybrid solar photovoltaic/regenerative fuel cell auxiliary power system (APS). This system was designed to power the primary control vehicle. The APS supplies life support power for four astronauts and a limited degree of mobility allowing the primary control vehicle to limp back to either a permanent base or an accent vehicle. The results showed that the APS described above, with a mass of 667 kg, was sufficient to provide live support power and a top speed of five km/h for 6 hours per day. It was also seen that the factors that had the largest effect on the APS mass were the life support power, the number of astronauts, and the PV cell efficiency. The topics covered include: (1) power system options; (2) rover layout and design; (3) parametric analysis of total mass and power requirements for a manned Mars rover; (4) radiation shield design; and (5) energy conversion systems.

El-Genk, Mohamed S.↗

NASA NTRS1991

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

El-Genk, Mohamed S.↗

NASA NTRS1991

Shrinkage void formation and its effect on freeze and thaw processes of lithium and lithium-fluoride for space applications

The effects of shrinkage void forming during freezing of lithium and lithium fluoride on subsequent thaw processes are investigated using a numerical scheme that is based on a single (solid/liquid) cell approach. Results show that a void forming at the wall appreciably reduces the solid-liquid interface velocity, during both freeze and thaw, and causes a substantial rise in the wall temperature during thaw. However, in the case of Li, the maximum wall temperature was much lower than the melting temperature of PWC-11, which is used as the structure material in the SP-100 system. Hence, it is concluded that a formation of hot spots is unlikely during the startup or restart of the SP-100 system.

Yang, Jae Y.↗

NASA NTRS1991

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

El-Genk, Mohamed S.↗

NASA NTRS1991

Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

Morley, Nicholas J.↗