SEARCH · Engineering Papers

Results for “liquid cooling”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 145 records · Page 8

All sources NASA NTRS DOE OSTI

DOE OSTI2023

Large Eddy Simulation of Convective Heat Transfer in a Random Pebble Bed Using the Spectral Element Method

The development of fluoride-cooled high-temperature reactors has drastically increased the demand for an in-depth understanding of the heat transfer (HT) in packed beds cooled by liquid salts. The complex flow fields and space-dependent porosity found in a pebble bed require a detailed understanding to ensure the proper cooling of the reactor core during normal and accident conditions. As detailed experimental data are complicated to obtain for these configurations, high-fidelity simulation such as large eddy simulation and direct numerical simulation (DNS) can be employed to create a high-resolution heat transfer numerical database that can assist in addressing industrial-driven issues associated with the heat transfer behavior of fluoride-cooled high-temperature reactors. In this paper, we performed a series of large eddy simulation using computational fluid dynamics (CFD) code NekRS to investigate the heat transfer for a bed of 1741 pebbles. Further, the characteristics of the flow, such as average, rms, and time series of velocity and temperature, have been analyzed. Porous media averages have also been performed. The simulation results show a good agreement between non-conjugate heat transfer and conjugate heat transfer. The generated data will be used to benchmark heat transfer modeling methods and local maxima/minima of heat transfer parameters. It will also be used for supporting convective heat transfer quantification for Kairos Power and benchmarking lower fidelity models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Ultra efficient turbo-compression cooling systems

Aspects of the present disclosure include a system for turbo-compression cooling. The system may be aboard a marine vessel. The system includes a power cycle and a cooling cycle. The power cycle includes a first working fluid, a waste heat boiler configured to evaporate the working fluid, a turbine, and a condenser. The condenser condenses the working fluid to a saturated or subcooled liquid. The cooling cycle includes a second working fluid, a first compressor configured to increase the pressure of the second working fluid, a condenser configured to condense the second working fluid to a saturated or subcooled liquid after exiting the first compressor, an expansion valve, and an evaporator. The turbine and first compressor are coupled one to the other. The waste heat boiler receives waste heat from engine jacket water and lubricating oil from a ship service generator. The evaporator cools water in a shipboard cooling loop.

Bandhauer, Todd M.↗

DOE OSTI2025

Structural Design and Modeling of MARVEL Primary Coolant System Using the ASME Section III, Division 5, Code

This paper presents the structural design and supporting analysis for the Microreactor Applications Research Validation and Evaluation (MARVEL) primary coolant system (PCS) using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Division 5, rules. MARVEL is a liquid metal–cooled microreactor intended to provide experimental capabilities for the rapid testing and development of microreactor technologies. The PCS utilizes high-temperature sodium-potassium liquid metal as the primary coolant and operates at a design temperature of 570°C, necessitating the consideration of creep-related failure mechanisms. The base metal for the PCS is 316H stainless steel, and the weldments are made with a 16-8-2 filler. The design approach incorporates the current base code rules along with ASME code cases N-924, N-861, and N-862 to address primary load, ratcheting, and creep-fatigue evaluations, respectively. The reactor’s operation involves complex thermal and mechanical interactions due to natural convective flow and differential thermal expansion between components. In conclusion, this paper discusses the structural engineering challenges encountered, such as managing thermal stresses in the distribution plenum and guard vessel, and outlines the strategies implemented to meet the code requirements, including design modifications and operational constraints.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2022

NREL Cooling Evaluation Capabilities

Panel presentation providing an overview of the Energy Systems Integration Facility (ESIF). Provided examples of thermal and power testing of computer components. Showcased the ESIF data center and testing capabilities for computer equipment at scale.

center↗

DOE OSTI2020

Combined condensing direct-contact exchange and separations

Devices, systems, and methods for separating a vapor from a gas are disclosed. A gas is passed through a direct-contact exchanger. The exchanger using a contact liquid to cool the gas. The gas comprises a vapor. A portion of the vapor is condensed as the gas passes through the direct-contact exchanger, producing a product liquid and a vapor-depleted gas. The product liquid is immiscible in the contact liquid. The product liquid is gravity settled from the contact liquid such that the contact liquid and the product liquid separate in the direct-contact exchanger.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

New High-Strength Ni-based Alloys for High Temperature Service in Liquid Fluoride Salt Environments

The ever-increasing demand for higher system thermal efficiency necessitates the operation of power generation cycles and heat conversion systems for chemical processes at progressively higher temperatures. As the system operating temperature increases, fewer and fewer materials are available with the required mechanical properties and environmental compatibility. This dearth of materials is particularly acute in structural applications at temperatures above 700°C in liquid Fluoride Salt Cooled High-Temperature Reactors (FHRs) and Concentrated Solar Power (CSP) systems where liquid fluoride and/or liquid chloride salts are used for their high thermal capacity and low thermal conductivity to store and transport heat. It has been shown that fluorides and chlorides can degrade the properties of the material used for storage and transport due to corrosive effects thus affecting the performance and lifetime characteristics. Thus, materials used for storage and transport for fluoride/chloride-salt must at the minimum have good compatibility with the salt by exhibiting low corrosion rates at the required temperatures. In addition, since many of the components will be subject to stresses in service, they need to have sufficient strength to resist plastic deformation, and the resistance to time dependent creep deformation at these temperatures, and stress levels. Achieving the combination of these properties required for higher temperatures is particularly challenging considering traditionally, the expected lifetime of some of the components used in these systems are from 30 years (CSP) to 80 years (FHRs). Although 316 stainless steel and associated redox corrosion reduction techniques are being considered for use in the first generational systems, use of 316SS is limited to operating temperatures up to ~ 650°C, thereby resulting in lower reactor efficiencies. Hastelloy ® N, developed at ORNL in the 1950s-60s for molten salt service is currently the leading candidate FHR structural alloy but has inadequate creep properties at high temperatures. Using a computationally-guided approach, Oak Ridge National Laboratory had developed new high strength alloys strengthened by γ’ precipitates (about 2X the strength of Hastelloy ® N) in small laboratory-scale heats that show good resistance to corrosion by fluorides and improved creep rupture life at temperatures up to 850°C (U.S. Patent Application No. 13/833,357 entitled “High Strength Alloys for High Temperature Service in Liquid-Salt Cooled Energy Systems”- granted U. S. Patent No. 9,540,714).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Disposition Options for Sodium Cooled Fast Reactor (A White Paper)

The sodium-cooled fast reactor (SFR) design concept is one of the six classes of nuclear reactors in the GenIV initiative. SFRs are uranium or plutonium-fueled reactors operating in the fast neutron spectrum using liquid sodium as the coolant. SFRs can be designed as a breeder reactor or actinide-burning reactor in addition to operating the thorium fuel cycle. While having different fuel designs, the anticipated waste streams, and the necessary management strategies for spent nuclear fuel (SNF) and radioactive wastes from SFRs are very similar. This includes the SNF, activated sodium coolant, in-core stainless-steel components, piping, resins and filters, solidified liquid waste, contaminated equipment, and other radioactive wastes. Modern SFR designs are based on a long and rich operating history of several liquid-metal-cooled fast reactors with sodium coolant. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the SNF and radioactive wastes associated with operating these reactors. This white paper will identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste and identify the key radioactive waste streams from SFRs. Idaho National Laboratory has significant experience in the management of the SNF from the SFR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new SFR-type small modular reactor or microreactor intended for deployment at the Idaho National Laboratory Site

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Cooling Design and Thermal Analysis for Thermal Shields of a Cryocooler-Cooled Superconducting ECR Ion Source MARS-D Magnet

Here, a demonstrative NbTi based Mixed Axial and Radial field System (MARS-D) is being developed for a next-Generation Electron Cyclotron Resonance Ion Source (ECRIS) at Lawrence Berkeley National Laboratory (LBL), which employs a novel closed-loop coil design scheme that more efficiently utilizes conductor fields and extend the application of NbTi for high frequency (up to 45 GHz) ECR operation. The NbTi MARS-D magnet consists of a single hexagonally shaped closed-loop coil and a set of auxiliary solenoids. A cryostat for cooling the MARS-D magnet is under design at LBL. The MARS-D magnet working around 4.2 K will be bath-cooled in liquid helium using multiple two-stage cryocoolers. An intermediate temperature thermal radiation shield is adopted to reduce the heat leakage imposed on 4.2 K coil cold mass from room temperature. The thermal shield is conduction-cooled by the first-stage cold heads of four two-stage cryocoolers and the cold head of a single-stage cryocooler shared with nine binary leads. The temperature in the area of the shield that warm ends of HTS leads are mounted on is expected no higher than 60 K, which is limited by maximum allowable working temperature of HTS leads. The paper presents thermal analysis on the thermal radiation shield including heat loads and effects of eddy current induced during quench on its material selection.

ECRIS↗

DOE OSTI2022

SAS4A/SASSYS-1 Validation with the FFTF LOFWOS Tests #10-12

The FFTF liquid sodium cooled test reactor was built to assist the development and testing of advanced fuels and material for fast reactors. The goals of the FFTF Passive Safety Testing program included confirming the safety margins of FFTF as a liquid metal reactor, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of its specific design features. The program included thirteen unprotected loss of flow without scram tests. The results of modeling and simulation of LOFWOS Tests #10, #11, and #12 are presented in this report to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. These three tests did not use the primary loop pony motors and thus transitioned to natural circulation flow rates from varying initial power levels. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project (CRP) for benchmark analysis of the FFTF LOFWOS Test #13. The SAS4A/SASSYS-1 model originally developed for the CRP has been extended to include LOFWOS Tests #10-12 and modernized to utilize new code features and modeling practices. Predicted results from SAS4A/SASSYS-1 were compared to measured test data and differences were investigated. Overall, it was concluded that, for LOFWOS Tests #10-12, there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

SAS4A/SASSYS-1 Validation with FFTF Tests

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Conjugate Multimode Heat Transfer Analysis of Cryogenic EXLO Manipulation

Abstract In this study, a conjugate radiation/conduction multimode heat transfer analysis of cryogenic focused ion beam (FIB) milling steps necessary for producing ex situ lift out specimens under cryogenic conditions (cryo-EXLO) is performed. Using finite volume for transient heat conduction and enclosure theory for radiation heat transfer, the analysis shows that as long as the specimen is attached or touching the FIB side wall trenches, the specimen will remain vitreous indefinitely, while actively cooled at liquid nitrogen (LN2) temperatures. To simulate the time needed to perform a transfer step to move the bulk sample containing the FIB-thinned specimen from the cryo-FIB to the cryo-EXLO cryostat, the LN2 temperature active cooling is turned off after steady-state conditions are reached and the specimen is monitored over time until the critical devitrification temperature is reached. Under these conditions, the sample will remain vitreous for >3 min, which is more than enough time needed to perform the cryo-transfer step from the FIB to the cryostat, which takes only ∼10 s. Cryo-transmission electron microscopy images of a manipulated cryo-EXLO yeast specimen prepared with cryo-FIB corroborates the heat transfer analysis.

Materials Science↗

DOE OSTI2021

Evaporation and propagation of liquid drop streams at vacuum pressures: Experiments and modeling

Evaporation of streams of liquid droplets in environments at vacuum pressures below the vapor pressure has not been widely studied. Here, experiments and simulations are reported that quantify the change in droplet diameter when a steady stream of ≈100 μm diameter drops is injected into a chamber initially evacuated to <10 –8 bar. In experiments, droplets fall through the center of a 0.8 m long liquid nitrogen cooled shroud, simulating infinity radiation and vapor mass flux boundary conditions. Furthermore, experimentally measured changes in drop diameters vary from ≈0 to 6 μm when the initial vapor pressure is increased from 10 –6 to 10 –3 bar by heating the liquid. Measured diameter changes are predicted by a model based on the Hertz-Knudsen equation. One uncertainty in the calculation is the “sticking coefficient” β. Assuming a constant β for all conditions studied here, predicted diameter changes best match measurements with β≈0.3. This value falls within the range of β reported in the literature for organic liquids. Finally, at the higher vapor pressure conditions considered here, the drop stream disperses transverse to the main flow direction. This spread is attributed to forces imparted by an absolute pressure gradient produced by the evaporating stream.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2022

An Innovative Zero-Liquid Discharge Intermediate-Cold-Liquid Eutectic-Freeze Desalination System

A novel freeze desalination technology is presented for purification of water produced from various industrial processes, including oil and gas extraction. The main innovative idea is introducing a water-immiscible intermediate-cold-liquid (ICL) medium which absorbs the cold from a refrigeration cycle and transfers it to a freezing chamber where it is mixed with brine. The cold ICL enters the freezing chamber at temperatures between -10°C to -30°C. Precooled brine in injected into the freezing chamber and pure water freezes out of the brine upon mixing with the ICL. The rate of freezing inside the chamber is controlled by adjusting the flow rate of the cold ICL relative to the inlet brine. Both eutectic freeze and non-eutectic freezing are possible by controlling the temperature of the freezing chamber. The solid-liquid slurry created in the freezing chamber flows to a filter separation unit after passing through a separation column. The ice crystals are separated from the liquid by a rotary filter device. The collected ice is rinsed by a centrifugal unit. The rinsed ice is melted to produce fresh water. The developed technology resolves the issues common to current freeze-desalination systems while offering superior heat transfer performance ascribed to direct-contact between the cooling medium and the brine in addition to producing solid salt and pure water. The system operates under atmospheric pressure. The low operation temperature minimizes the corrosion issues common in systems exposed to brine. Also, due to the low operating temperature, there are no concerns about evaporation of volatile organics from the produced water during the treatment. The developed technology effectively eliminates the use of chemicals for pretreatment of brine. The system can be constructed from commercially available components. Extensive testing using a lab-scale prototype by using synthetic brine and actual produced water with TDS values from 50,000 ppm to 300,000 ppm showed that freshwater could be achieved with relatively small dependency on the quality of the input brine. The quality of the treated water was found to be mainly a function of the duration of the centrifugal process. Treated water with TDS as low as 200 ppm could be achieved in the tests. Thermo-economic modeling coupled with system level thermochemical process model were developed for prediction of the LCOW. The LCOW was predicted to be $\$$0.50 to $\$$0.90 per barrel of produced water. The potential markets for deployment of the technology are the oil and gas, in-land thermoelectric powerplants, RO desalination plants, and mining industry.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2021

History and status of spent fuel treatment at the INL Fuel Conditioning Facility

Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Benchmark Specifications for Select Experiments Conducted at the Kansas State University Gallium Thermal-hydraulic Experiment Facility

The Department of Energy (DOE) – Nuclear Energy University Programs (NEUP) supported the creation and operation of the Gallium Thermal-hydraulic Experiment (GaTE) facility at Kansas State University (KSU) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. GaTE was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements describing stratification behavior in the coolant were collected for use as experimental benchmarks in validation efforts for computational models. The results of these tests contribute to a greater understanding of thermal stratification behavior of liquid metal under various configurations and operating conditions. This report provides a complete description of the benchmark problem, including all necessary details and description of a set of four forced flow and four natural circulation tests and measured data for comparison with model results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Benchmark Specification for Select Experiments Conducted at the University of Wisconsin-Madison Thermal Stratification Test Facility

The Department of Energy (DOE)-Nuclear Energy University Programs (NEUP) supported the creation and operation of the Thermal Stratification Test Facility (TSTF) at the University of Wisconsin Madison (UWM) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. The TSTF was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements of the flow and temperature were collected for use as experimental benchmarks to support validation efforts for computational models. The results of these tests contribute to the greater understanding of thermal stratification behavior of liquid sodium under various configurations and operating conditions. The six TSTF tests selected for benchmarking are a set of forced circulation tests at a fixed flow rate with different Upper Internal Structure (UIS) configurations in the test section (no UIS, solid UIS, and a UIS with flow area of 4, 8, 12, and 100%). This report provides a complete description of the benchmark problems, including all key test facility details, descriptions of each test condition, and measured data for comparison with modeled results.

42 ENGINEERING↗

DOE OSTI2025

A cryogenic white light absorption spectroscopy setup with in situ gamma irradiation and thermo-optical annealing for optical fiber radiation-induced attenuation characterization

Silica optical fiber sensors offer a fast, distributed measurement solution in various cryogenic and radiation environments, such as magnets for fusion power. Under these conditions, light-absorbing point defects limit lifetime via radiation-induced attenuation (RIA). To support RIA kinetics prediction, we present an in situ, broadband absorption spectroscopy setup combining gamma irradiation with liquid nitrogen cooling. The apparatus enables continuous monitoring of narrowband RIA levels and the use of secondary optical annealing light sources, alongside broadband spectrum measurements to characterize RIA defects through spectrum decomposition. Initial results confirm the inevitable photobleaching effect of the probe light source, which must be accounted for. In addition, we report RIA kinetics during cycles of gamma irradiation at 77 K and isochronal thermal annealing steps from liquid nitrogen to room temperature, showcasing the setup’s ability to replicate real fiber operation scenarios. These instruments form an ideal platform to further study the kinetics of RIA coupled with thermal and optical annealing.

Absorption spectroscopy↗

DOE OSTI2022

Influence of design and operating parameters for additively manufactured intensified packing devices on CO 2 -Absorption column cooling and capture efficiency

Solvent-based CO 2 absorption is technologically a matured CO 2 capture pathway but suffers from: high regeneration energy demand, and solvent temperature rise and decreased capture efficiency caused by the heat of reaction. While research has focused on developing non-aqueous and low-aqueous solvents for decreasing the regeneration energy, the temperature bulge due to exothermic absorption is typically dealt with by cooling the solvent with an external inter-stage heat-exchanger. This approach may increase the overall process footprint, as well as capital and operating costs. Here, the current study explores a process intensification approach by incorporating inside the column an additively manufactured intensified packing device that consists of corrugated plates and internal coolant channels. The corrugated plates provide surface area for mass transfer between gas and liquid, while cooling fluid inside the internal channels removes heat from the exothermic reactive system. Two different intensified devices with specific surface areas of 266 m 2 /m 3 and 359 m 2 /m 3 were designed, manufactured, and tested inside a 2.06-m long and 0.203-m diameter column packed with commercial Mellapak 250Y packing. The device with a lower surface area showed up to 27 % reduction in cooling performance. A steady-state heat transfer model provides good agreement with the experimental column temperature and intra-stage heat removal data. Although a decrease in heat transfer performance was observed for the device with lower surface area, CO 2 capture experiments performed with simulated flue gas and low-aqueous solvent demonstrated that both intensified devices lead to a similar improvement of 12 % in capture efficiency. This study provides an understanding on simplifying the intensified packing device geometry and decreasing the device fabrication cost by 25 % without compromising with the CO 2 absorption performance of the packed column.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗