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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Design, Fabrication, and testing of Direct Air Capture Sorbent modules to increase understanding of trade-offs in pressure drop and capture efficiency

Conference presentation for the work entitled “Design, Fabrication, and testing of Direct Air Capture Sorbent modules to increase understanding of trade-offs in pressure drop and capture efficiency”. In recent years, there has been a growing interest in Direct Air Capture (DAC) of CO2. This technology is gaining attention as a method to enhance the reduction of atmospheric greenhouse gases (GHGs), particularly when used alongside other decarbonization strategies. The route for efficient and net carbon negative DAC is challenging because CO2 in the atmosphere is highly diluted (~400ppm). This work, utilizing 3D printing capabilities for rapid prototyping, presents application-based findings on the design, fabrication, and testing of DAC module form factors that house solid sorbents. The work leverages an experimental rig that can achieve air throughput and velocities approaching that which are anticipated for commercial scales. The tradeoffs of pressure drop, footprint and capture efficiency are discussed for different orientations of a NETL patented sorbent material.

Riley, Jarrett↗
DOE OSTI2021

Introduction of the PELICAN loop, a Full-Scale Pressure Drop Test Facility

The Versatile Test Reactor (VTR) is a test reactor currently under development by the US Department of Energy. This reactor will rely on fast neutrons enabling novel and wide-ranging experiment to support the development of the various advanced reactor technologies. With the high flux achievable, accelerated testing of fluid and materials will be made possible. To support VTR design efforts [1], an experimental facility has been designed and constructed at Argonne National Laboratory to recreate the hydraulic flow conditions within the VTR’s primary heat transport system (PHTS). This facility, the Pressure drop Experimental Loop for Investigations of Core Assemblies in advanced Nuclear reactors, PELICAN, measures the pressure drop across a full-scale fuel assembly containing prototypic axial reflectors, fuel, and plena components. Here, we first describe the design considerations required to recreate aspects of the VTR. Then we discuss the design and construction of PELICAN to address these design requirements, the design and construction of the test articles placed inside PELICAN’s test section, and finally present some of the first experimental results.

Grannan, A. M.↗
DOE OSTI2021

Heat Transfer and Pressure Drop Characteristics of Water Flow Boiling in Internally Enhanced Tubes

Flow boiling is a critical process for many thermal conversion processes such as HVAC&R, water heating, power generation and desalination applications. The major goal of this paper is to enhance the in-tube heat transfer for flow boiling of water and further improve the efficiency of the boiler. This paper presents an experimental investigation of the heat transfer and pressure drop characteristics for water under forced convection and boiling conditions inside both smooth and enhanced tubes. The results in the smooth copper tube are compared with several correlations and used as a baseline. The two-phase heat transfer and pressure drop performance of the water flow boiling in the internally enhanced tube is investigated, and the influence of internal surface pattern on the thermal characteristics is discussed. The relations of the heat transfer coefficient with mass flux, heat flux and vapor quality are analyzed according to the results.

Yang, Cheng-Min↗
DOE OSTI2022

Pressure Drop Correlation Improvement for the Near-Wall Region of Pebble-Bed Reactors

Packed beds play an important role in several engineering fields, with their applications in nuclear energy being driven by the development of next-generation reactors utilizing pebble fuel. The random nature of a packed pebble bed creates a flow field that is complex and difficult to predict. Porous media models are an attractive option for modeling pebble-bed reactors (PBRs), as they provide intermediate fidelity results and are computationally efficient. Porous media models, however, rely on the use of correlations to estimate the effect of complicated flow features on the pressure drop and heat transfer in the system. Existing correlations were developed to predict the average behavior of the bed, but they are inaccurate in the near-wall region where the presence of the wall affects the pebble packing. This work aims to investigate the accuracy of a porous media model using the Kerntechnischer Ausschuss (KTA) correlation, the most common pressure drop correlation for PBRs compared to the high-fidelity large eddy simulation (LES). A bed of 1568 pebbles is investigated at Reynolds numbers from 625 to 10 000. The bed is divided into five concentric subdomains to compare the average velocity, friction losses, and form losses between the porous media and LES codes. The comparison between the LES simulation and the KTA correlation revealed that the KTA correlation largely underpredicts the form losses in the near-wall region, leading to an overprediction of the velocity near the wall by nearly 30%. An investigation of the form losses across the range of Reynolds numbers in the LES results provided additional insight into how the KTA correlation may be improved to better predict these spatial effects in a pebble bed. These data suggest that the form coefficient near the wall must be increased by 48% while decreasing the form coefficient of the inner bulk region of the bed by 15%. The implementation of these improvements to the KTA correlation in a porous media model produced a radial velocity profile that saw significantly improved agreement with the LES results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Condensation heat transfer and pressure drop of low-global warming potential refrigerants in smooth aluminum tubes

In this work we report condensation heat transfer coefficient and pressure drop data for R-454B, R-32, and R-454C. These refrigerants are low Global Warming Potential (GWP) replacements for R-410A. We obtained these data in a smooth 7.2 mm outer diameter (5.94 mm inner diameter) aluminum tube. In these experiments, the key variables were condensation temperature (35 °C ≤ T cond ≤ 50 °C) and refrigerant mass flux (151.1 ≤ G ≤ 347.3 kg m –2 s –1 ). In general, for similar nominal operating conditions, we measured the highest heat transfer coefficients for R-32 and the lowest for R-454C. Among the correlations we used to predict the heat transfer data for all three refrigerants, the one developed by Thome et al. (2003) predicted the data with the greatest accuracy. Deviations between experimental data and predictions of this correlation were 21.7% for R-454B (with SBG correction), 22.0% for R-32, and 18.8% for R-454C (with SBG correction). Experimental pressure gradient data were predicted by the Friedel correlation Friedel (1979) with deviations of 13.8% for R-454B, 9.8% for R-32, and 10.2% for R-454C. In this work, we also compare the predictions of the Cavallini et al. correlation Cavallini et al. (2006) for R-454B and R-32 in smooth copper versus smooth aluminum tubes. To improve this correlation’s predictive capability for aluminum tubes, we have proposed some modifications. After incorporating these modifications, the deviations between the correlation’s predictions and our data were 10% for R-454B (with SBG correction), 8.1% for R-32, and 5.2% for R-454C (with SBG correction). Furthermore, these data and findings will assist the HVAC and refrigeration industry to implement low-GWP refrigerants in condensers using smooth aluminum tubes.

Aluminum tube↗
DOE OSTI2021

Pronghorn Porous Media Model Validation with Pressure Drop Measurements

The verification and validation (V &V) of Pronghorn is imperative to assert its accuracy when predicting the fluid velocity, temperature, and pressure in high temperature gas-cooled reactors. This paper focuses on the validation of the Pronghorn implementation of the incompressible and compressible Navier-Stokes equations using the pressure drop measurements in the engineering-scale pebble bed facility at the Texas A &M university.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Pressure drop and heat transfer characteristics of nitrate salt and supercritical CO 2 in a diffusion-bonded heat exchanger

A lab-scale 316/L stainless steel diffusion-bonded heat exchanger was examined within a coupled high-temperature test facility consisting of an atmospheric-pressure nitrate salt loop and a supercritical carbon dioxide (sCO 2 ) loop operating between 10 and 16 MPa. Pressure loss measurements collected for flow rates up to 0.55 kg/s of Solar Salt and up to 0.6 kg/s of carbon dioxide were used to generate friction factor correlations for the respective circular and semicircular zig-zag flow passages. Heat transfer measurements were also performed over a wide range of conditions to produce Nusselt correlations for the CO 2 and nitrate salt geometries. The Nusselt number is deduced from a heat transfer resistance network and overall conductance measurements collected across the operating envelope of the coupled sCO 2 – nitrate salt test facility. The CO 2 side correlation showed excellent agreement with existing correlations in the literature. Further, the salt side, having nearly circular channels with a different zig-zag angle and hydraulic diameter, yielded a new correlation applicable to the laminar flow regime. To examine the similitude of the friction correlations, pressure drop testing with water was performed on both sides of the component as well. Finally, results from non-destructive examination are discussed to provide insights regarding inspection methods and draining performance of high-temperature liquids in compact heat exchangers.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor

This work studies the hydrodynamics of the seven pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. Here, the scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number Re. A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of Re values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors f Exp agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within ±7% but are significantly underpredicted by Rehme's correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall y + values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method (y + ~ 1) with the Shear Stress Transport k - ω turbulence model and the scalable wall functions (y + > 12) with the realizable k - ε turbulence model can predict f Exp within ±5% The standard wall functions (y + > 30) and nonequilibrium wall functions (y + > 30) with the realizable k - ε model can predict f Exp within ±10%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

CFD Benchmark of Pressure Drop in a 61-Pin Wire-Wrapped Assembly with Blocked Channels Using NekRS

Thermal hydraulic behaviors of wire-wrapped rod bundles continue to be a subject of intense research. One of the leading next-generation designs, the sodium fast reactor, relies on a hexagonal assembly of wire-wrapped fuel pins. The issue of potential channel blockage has been raised as a safety concern due to the tightly packed arrangement of the fuel pins. This has led to several recent experimental and computational studies working to quantify the potential impact on the fluid flow and heat transfer behaviors of such blockages. The objective of the present study is to benchmark the high-fidelity NekRS CFD solver in predicting pressure drop for large blockages against available experimental data. A 61-pin wire-wrapped fuel assembly with two flow blockage configurations has been simulated and investigated at various low to moderate Reynolds numbers. The NekRS solver has been shown to yield exponentially decreasing spatial discretization errors with increasing polynomial order. All simulated results agreed well with measured data, which indicates that the overall methodology is adequate and consistent. The results of this benchmark study demonstrate the accuracy of NekRS for sodium fast reactor hydrodynamic simulations, increasing the confidence in its use for design, licensing, and analysis activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Low pressure drop advanced swirl technology gas filter

A media-free filter device for a supercritical fluid process is disclosed. The media-free filter device includes a first pipe section connected in flow communication with a closed-loop system for a supercritical fluid. A y-pipe segment intersects the first pipe section at an intersection joint. The intersection joint intersects at a descending obtuse angle with respect to the first pipe section flow direction. The y-pipe segment is open at the intersection joint and closed opposite the intersection joint, in flow communication with the first pipe section. The y-pipe section induces an eddy current turbulence zone adjacent the intersection joint to remove substantially all contaminants suspended in the supercritical fluid; and the contaminants are deposited in the descending y-pipe segment. An SCO2 process incorporating the media-free filter device is also disclosed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗