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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Enhanced Models for the Simulation of Two-Phase Boiling Flow and Critical Heat Flux with the STAR-CCM+ and NEK-2P CFD Codes: Executive Summary (Final Report)

The goal of the project was to develop, verify and validate an advanced two-phase boiling flow modeling capability that enhances the existing two-phase flow modeling of the STAR-CCM+ Computational Fluid Dynamics (CFD) code. The project focused on the simulation of two phase flow and heat transfer phenomena that occur in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) fuel bundles under various operating conditions. The extended STAR-CCM+ two-phase flow models implemented during the project follow the approach similar to the Extended Boiling Framework [1-6] previously developed at Argonne but include more fundamental physical models of boiling phenomena. The development and validation of the extended two-phase models is supported by the U.S. DOE GAIN program and performed in collaboration with Framatome U.S. Inc.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1962

STABILITY OF FILM BOILING TWO-PHASE FLOW IN CRYOGENIC SYSTEMS

The stability of film boiling two phase flow is investigated referring in particular to saturated cryogenic liquids. This type of fluid motion is unstable, such that a range of wave numbers will always be present in which disturbances are amplified. Further the mode of operation of the energy supply system may favour temperature fluctuations of large amplitude if the circuit does not provide feed-back compensation. The theoretically predicted stabilization due to capillarity is reduced considerably in liquefied gases with low surface tension, such as hydrogen. This effect causes the transition to turbulent flow to occur at low Reynolds numbers, and increases the heat transport in the turbulent regime. A comparison with experimental data, (horizontal cylinder), shows that the "critical" film Reynolds number ūδ/𝜈, at which instabilities become effective, is lowered from approximately 200, (O2), to about 30, (H2). Similar results have been found with other configurations. Therefore application of laminar film boiling solutions is restricted to small objects, entrance regions, or stagnation points. Turbulent correlations given so far are valid only for a certain group of substances with similar behavior.

LIQUID FLOW↗
NASA NTRS1994

Bubble dynamics, two-phase flow, and boiling heat transfer in a microgravity environment

The two-phase bubbly flow and boiling heat transfer in microgravity represents a substantial challenge to scientists and engineers and yet there is an urgent need to seek fundamental understanding in this area for future spacecraft design and space missions. At Washington State University, we have successfully designed, built and tested a 2.1 second drop tower with an innovation airbag deceleration system. Microgravity boiling experiments performed in our 0.6 second Drop Tower produced data flow visualizations that agree with published results and also provide some new understanding concerning flow boiling and microgravity bubble behavior. On the analytical and numerical work, the edge effects of finite divergent electrode plates on the forces experienced by bubbles were investigated. Boiling in a concentric cylinder microgravity and an electric field was numerically predicted. We also completed a feasibility study for microgravity boiling in an acoustic field.

Chung, Jacob N.↗
NASA NTRS1993

The bubbly-slug transition in a boiling two-phase flow under microgravity

A theory is presented to describe, in reduced gravity flow boiling, the transition from bubbly two-phase flow to slug flow. It is shown that characteristics of the bubbly flow and the transition were controlled by the mechanism of vapor bubble growth dynamics. By considering in nucleate boiling, behavior of vapor bubbles at departure from a heated surface a condition required for transition was determined. Although required, this condition alone could not ensure coalescence of bubbles to cause the transition to slug two-phase flow. The condition leading to coalescence, therefore, was obtained by examining oscillations of vapor bubbles following their departure from the heated surface. The predicted transition conditions were compared with the prediction and test data reported for adiabatic reduced gravity two-phase flow, and good qualitative agreement was found.

Kiper, Ali M.↗
DOE OSTI2019

Impact of dissolved salts on two-phase flow and boiling heat transfer in a natural circulation loop

Two-phase natural circulation and boiling experiments are conducted in this study with artificial seawater as a heat transfer fluid or process coolant. This experimental work is focused on understanding the role of dissolved salts in water on the steady state two-phase natural circulation flow rates. Mass flow rates were measured at different heat flux levels with artificial seawater, deionized water and tap water in the experimental loop. It was found that the average quasi-steady state mass flow rates were the same at any particular heat flux level regardless of the fluid being used. Artificial seawater coolant exhibited more stable natural circulation at higher heat flux as compared to the other two coolants. Mass flow measurements recorded at 2 s intervals show that magnitude of flow oscillations are much higher in case of tap water and deionized water. Also, high speed camera images revealed that the bubble departure diameters are smaller in seawater compared to tap or deionized water. This can be attributed to the increased local wettability due to the presence of dissolved salts in the fluid. Visual observations of the riser section showed that bubble coalescence and formation of slugs is inhibited in the seawater. These observations also explain the reduced pressure drop across test section when seawater was used as a coolant.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

X-ray Imaging-Based Void Fraction Measurement in Saturated Flow Boiling Experiments with Seawater Coolant

Recent studies have shown that the presence of dissolved salts in water can exhibit peculiar flow boiling and two-phase flow regimes. Two-phase flow and convective flow boiling are typically characterized with the help of void fraction measurements. To quantitatively improve our understanding of two-phase flow and boiling phenomenon with seawater coolant, void fraction data are needed, which can not be obtained from optical imaging. In this paper, we present experimental void fraction measurements of saturated flow boiling of tap water and seawater using X-ray radiography. X-rays with a maximum energy level of 40 KeV were used for imaging the exit region of the heated test section. At lower heat flux levels, the two phase flow in seawater was bubbly and homogeneous in nature, resulting in higher void fractions as compared to tap water. With an increase in heat flux, the flow regime was similar to slug flow, and void fraction measurements approached similarity with tap water. The predicted pressure drop using the measured void faction shows good agreement with the measured total pressure drop across the test section, demonstrating the validity of the measurement process.

Eckels, Seth↗
DOE OSTI2016

Time-resolved measurement of subchannel void distribution for boiling two-phase flow in 5 x 5 rod bundle at high-pressure and high-temperature

A subchannel void sensor (SCVS) was improved to acquire time-resolve boiling two-phase flow in a 5-by-5 heated rod bundle at high-pressure and high-temperature. The SCVS consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes. The SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region in addition to 100 points of rod surface regions between rods. The devised sensors are installed in the heated rod bundle at eight height levels to acquire two-phase flow dynamics and its development. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The paper addresses the axial and cross-sectional distributions of void fraction at various pressure and rod bundle power conditions, which simulates an accidental condition in reactor core and in a spent fuel pool of a boiling water reactor (BWR). The experimental data are suitable to validate the computational multi-fluid dynamics (CMFD) and subchannel codes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1998

Bubble Dynamics, Two-Phase Flow, and Boiling Heat Transfer in Microgravity

This report contains two independent sections. Part one is titled "Terrestrial and Microgravity Pool Boiling Heat Transfer and Critical heat flux phenomenon in an acoustic standing wave." Terrestrial and microgravity pool boiling heat transfer experiments were performed in the presence of a standing acoustic wave from a platinum wire resistance heater using degassed FC-72 Fluorinert liquid. The sound wave was created by driving a half wavelength resonator at a frequency of 10.15 kHz. Microgravity conditions were created using the 2.1 second drop tower on the campus of Washington State University. Burnout of the heater wire, often encountered with heat flux controlled systems, was avoided by using a constant temperature controller to regulate the heater wire temperature. The amplitude of the acoustic standing wave was increased from 28 kPa to over 70 kPa and these pressure measurements were made using a hydrophone fabricated with a small piezoelectric ceramic. Cavitation incurred during experiments at higher acoustic amplitudes contributed to the vapor bubble dynamics and heat transfer. The heater wire was positioned at three different locations within the acoustic field: the acoustic node, antinode, and halfway between these locations. Complete boiling curves are presented to show how the applied acoustic field enhanced boiling heat transfer and increased critical heat flux in microgravity and terrestrial environments. Video images provide information on the interaction between the vapor bubbles and the acoustic field. Part two is titled, "Design and qualification of a microscale heater array for use in boiling heat transfer." This part is summarized herein. Boiling heat transfer is an efficient means of heat transfer because a large amount of heat can be removed from a surface using a relatively small temperature difference between the surface and the bulk liquid. However, the mechanisms that govern boiling heat transfer are not well understood. Measurements of wall temperature and heat flux near the wall would add to the database of knowledge which is necessary to understand the mechanisms of nucleate boiling. A heater array has been developed which contains 96 heater elements within a 2.5 mm square area. The temperature of each heater element is held constant by an electronic control system similar to a hot-wire anemometer. The voltage that is being applied to each heater element can be measured and digitized using a high-speed Analog to Digital (A/D) converter, and this digital information can be compiled into a series of heat-flux maps. Information for up to 10,000 heat flux maps can be obtained each second. The heater control system, the A/D system and the heater array construction are described in detail. Results are presented which show that this is an effective method of measuring the local heat flux during nucleate and transition boiling. Heat flux maps are obtained for pool boiling in FC-72 on a horizontal surface. Local heat flux variations are shown to be three to six times larger than variations in the spatially averaged heat flux.

Chung, Jacob N.↗
DOE OSTI2020

Development and Validation of a Two-Phase Thermal-Hydraulic CFD Code NEK-2P

A project is underway to develop, verify and validate an advanced two-phase flow modeling capability for the highly-scalable, high-performance Computational Fluid Dynamics (CFD) code NEK5000. The goal of this work is to verify and validate the two-phase version of the NEK5000 code, named NEK-2P, to simulate the two-phase flow and heat transfer phenomena that occur in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. The NEK-2P two-phase flow models follow the approach used for the Extended Boiling Framework (EBF) previously developed at Argonne but include more fundamental physical models of boiling phenomena and advanced numerical algorithms for improved computational accuracy, robustness, and computational speed. The development of the NEK-2P two-phase solver and the implementation of the Extended Boiling Framework two-phase models were initially supported by Argonne National Laboratory (Argonne) through a Laboratory Directed Research and Development (LDRD) project during FY2014-2016. The development and validation of the two-phase models through analyses of selected two-phase boiling flow experiments was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY2017-2020. This report focuses on verification and validation of the water-steam boiling model NEK-2P Two-Phase, CFD code. The NEK-2P was validated with Nuclear Power Engineering Corporation (NUPEC) Pressurized Water Reactor (PWR) Sub-channel and Bundle Test (PSBT) void distribution benchmark. Three different simulations were performed and analyzed for various operating conditions such as wall-heat flux and sub-cooled inlet temperatures. Reasonably good agreement with measured data was obtained in predicting the measured void distributions. Simulations were performed for Virginia Tech. (VT) 3x3 rod bundle geometry with and without spacers. The preliminary results were presented for Simplified Spacer Grid (SSG). In addition, the implementation of interface reconstruction model was tested with one of the Becker benchmark Critical Heat Flux (CHF) experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development and Validation of a Conjugate Heat Transfer Model for the Two-Phase CFD Code NEK-2P

A project is underway to develop, verify and validate an advanced two-phase flow modeling capability for the highly-scalable, high-performance CFD code NEK5000. The goal of the project is to develop a new two-phase version of the NEK5000 code, named NEK-2P, to simulate the two-phase flow and heat transfer phenomena that occur in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. The NEK-2P two-phase flow models follow the approach used for the Extended Boiling Framework (EBF) [2-3] previously developed at Argonne, but include more fundamental physical models of boiling phenomena and advanced numerical algorithms for improved computational accuracy, robustness, and computational speed. The development of the NEK-2P two-phase solver and the implementation of the Extended Boiling Framework two-phase models were initially supported by Argonne National Laboratory (Argonne) through a Laboratory Directed Research and Development (LDRD) project during FY2014-2016. The development and validation of the two-phase models through analyses of selected two-phase boiling flow experiments was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY2017-2019. This report focuses on the FY2019 development and validation of the Conjugate Heat Transfer (CHT) model for the NEK-2P Two-Phase, Computational Fluid Dynamics (CFD) code. The NEK-2P CHT model calculates the coupled behavior of the fluid and solid domains. In the fluid domain NEK-2P calculates the vapor and liquid phase temperatures, velocities and mass fractions. In the solid domain only the temperatures are calculated. The fluid and solid domains are coupled with a heat flux boundary condition at the fluid-solid interface. The heat flux and the wall temperatures at the fluid-solid interface are updated using the NEK-2P wall heat-flux partitioning model. A wall heat flux boundary condition is also applied to the exterior surface of the solid domain. The CHT model validation was illustrated through the analysis of three of the Becker Critical Heat Flux (CHF) tests. Reasonably good agreement with measured data was obtained in predicting the CHF location and post CHF wall temperature behavior illustrating the ability of the NEK-2P CHT model to simulate the CHF phenomena for a wide range of thermal-hydraulic conditions. Apart from the CHT model development and validation, initial simulations were performed for the Virginia Tech. (VT) 3x3 bare rod bundle experiments including both cold and boiling flow simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Simulation of Boiling Two-Phase Flow in a Helical Coil Steam Generator Using the Spectral Element Code Nek-2P

To enable the design of a light water small modular reactor, the boiling flow inside a helical coil steam generator has been simulated with the two-fluid model in Nek-2P. Nek-2P is the multiphase branch of the spectral element code Nek5000. Details of the implementation of the two-fluid model and the included closure models are discussed. The presented closure models include interactions for momentum, heat, and mass transfer between phases. Models for the drag, lift, and turbulent dispersion forces are included. The complete model is fully consistent in the limits of both phasic volume fractions approaching zero and is able to simulate flows of dispersed vapor, continuous liquid, dispersed liquid, continuous vapor, or any combination thereof. The closure models and their implementation in Nek-2P have been validated by comparing to experimental data for a boiling flow, demonstrating excellent agreement. Results from the simulation of the helical coil indicate strong phasic separation driven by the effects of buoyancy and inertia. Finally, significant differences were observed in the results compared to simulations performed using Star-CCM+, although these differences were somewhat expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2020

Case Study of Using Flocad to Model a Ground Test Station Ln2 Heat Exchanger

This paper presents a case study of using FLOCAD to model the heat transfer and boiling flow behavior of a Liquid Nitrogen (LN2) ground station equipment heat exchanger. The unique aspect of this work is that is one of very few studies documenting the use of FLOCAD in the spacecraft thermal control community. The results of the simulations described herein were used to guide and plan the execution of thermal vacuum testing spacecraft hardware using a ground test heat exchanger such as the one modeled herein. The paper will review the theory of two-phase boiling flow and correlations used by SINDA/FLUINT as well as the terminology and nomenclature used within Thermal Desktop / FLOCAD regarding two-phase boiling flow heat transfer modeling and simulation. Results for using the ground station heat exchanger as a heat sink in a thermal vacuum test set up are shown in order to demonstrate the set-up and execution of a typical FLOCAD model. Results indicate that properly modeling of the two-phase behavior is critical in order to ascertain the time constant associated with the heat exchanger, as well as understanding the temperature distribution across the test equipment and prediction of required LN2 flowrates to be used during testing. The paper is meant to augment the FLOCAD user’s manual and serve as a tutorial for thermal engineers and analysts wishing to learn and apply FLOCAD to industrial problems.

Anderson, Kevin R.↗
DOE OSTI2016

Investigation on numerical method of one-dimensional drift flux model for boiling two-phase flow

A numerical method for the solution of one-dimension drift-flux model has been developed. Correlations for heat and mass transfer and kinematic constitutive equation were used to close the set of the partial differential equations. The finite volume method based on the staggered grid and fully implicit backward Euler method were applied to discretize the spatial and time differential items, respectively. The discretization equations were solved iteratively in a same manner as in SIMPLE method for a single-phase flow. The verification and validation of the numerical method and closure model for one-dimension drift-flux model have been performed by comparing the code solution with the experimental data. Furthermore, the axial distribution of void fraction and pressure also have been investigated under different heated case. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗