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DOE OSTI2017

Verification of bubble tracking method and DNS examinations of single- and two-phase turbulent channel flows

Direct numerical simulation (DNS) has been regarded as a reliable data source for the development and validation of turbulence models along with experiments. The realization of DNS usually involves a very fine mesh that should be able to resolve all relevant turbulence scales down to Kolmogorov scale [1]. As the most computationally expensive approach compared to other CFD techniques, DNS applications used to be limited to flow studies at very low Reynolds numbers. Thanks to the tremendous growth of computing power over the past decades, the simulation capability of DNS has now started overlapping with some of the most challenging engineering problems. One of those examples in nuclear engineering is the turbulent coolant flow inside reactor cores. Coupled with interface tracking methods (ITM), the simulation capability of DNS can be extended to more complicated two-phase flow regimes. Departure from nucleate boiling (DNB) is the limiting critical heat flux phenomena for the majority of accidents that are postulated to occur in pressurized water reactors (PWR) [2]. As one of the major modeling and simulation (M&S) challenges pursued by CASL, the prediction capability is being developed for the onset of DNB utilizing multiphase-CFD (M-CFD) approach. DNS (coupled with ITM) can be employed to provide closure law information for the multiphase flow modeling at CFD scale. In the presented work, research groups at NCSU and UND will focus on applying different ITM to different geometries. Higher void fraction flow analysis at reactor prototypical conditions will be performed, and novel analysis methods will be developed, implemented and verified for the challenging flow conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS1980

Conceptual design of two-phase fluid mechanics and heat transfer facility for spacelab

Five specific experiments were analyzed to provide definition of experiments designed to evaluate two phase fluid behavior in low gravity. The conceptual design represents a fluid mechanics and heat transfer facility for a double rack in Spacelab. The five experiments are two phase flow patterns and pressure drop, flow boiling, liquid reorientation, and interface bubble dynamics. Hardware was sized, instrumentation and data recording requirements defined, and the five experiments were installed as an integrated experimental package. Applicable available hardware was selected in the experiment design and total experiment program costs were defined.

North, B. F.↗

DOE OSTI1988

Nuclear Safety [Vol. 29, No. 1, January-March 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. THE CHERNOBYL ACCIDENT: 1 Core History and Nuclide Inventory of the Chernobyl Core at the Time of Accident by Gerald Kirchner and Cornelius C. Noack; GENERAL SAFETY CONSIDERATIONS: 6 Nonprescriptive Nuclear Safety Regulation: The Example of Loss of Offsite Power by M. W. Golay, V. P. Manno, and C. Vlahoplus, Jr., 20 Nuclear Power Safety Goals in Light of the Chernobyl Accident by C. Whipple and C. Starr; ACCIDENT ANALYSIS: 29 Containment Loads from Severe Accidents—U.S. Program by W. Kerr and M. K. Dey; PLANT SAFETY FEATURES: 36 Safety Characteristics of Modern High-Temperature Reactors: Focus on German Designs by W. Kröger, H. Nickel, and R. Schulten; ENVIRONMENTAL EFFECTS: 49 Radiation Hormesis and Nuclear Safety by C. C. Congdon; OPERATING EXPERIENCES: 58 Evaluation of Nonradiological Water Chemistry at Power Reactors by Harvey Zibulsky, James J. Kottan, Walter J. Pasciak, Mary Ann Castrogivanni, and Sujit Banerjee, 64 Reactor Shutdown Experience Compiled by J. W. Cletcher, 67 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 85 General Administrative Activities Compiled by E. G. Silver, 94 Reports, Standards, and Safety Guides by D. S. Queener, 100 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 103 Proposed Rule Changes as of Sept. 30, 1987; ANNOUNCEMENTS: 5 Northwestern University Short Course on Radiation Safety, 28 International Approach to Nuclear Safety (After Three-Mile Island and Chernobyl), 57 RPI Short Course on Modern Developments in Boiling Heat Transfer and Two-Phase Flow, 107 20th DOE/NRC Nuclear Air Cleaning Conference, 110 Third International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, 108 The Authors, 111 Indexes to Nuclear Safety, Volume 28.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Implementation of a pressure correction term in the CTF sub-channel code

CTF [1] is a modernized and improved version of the legacy subchannel code, COBRA-TF [2], that is being jointly developed and maintained by Oak Ridge National Laboratory (ORNL) and North Carolina State University. The code was adopted by ORNL for use in the Consortium for Advanced Simulation of Light Water Reactors (CASL) in 2012 for aiding in addressing CASL challenge problems. Since that time, activities related to CTF have included implementing software quality assurance measures, implementing new features and models, performing validation and verification testing, establishing and supporting a CTF User Group, and developing the code for use in coupled applications [3, 4, 5]. CTF is applicable to single- and two-phase flows in light water reactor geometries at normal and accident operating conditions. The two-phase flow model consists of the six-equation two-phase flow model [6] augmented by equations for the droplet field and also includes appropriate source terms. This work focuses on the numerical stability of the six-equation two-phase flow model and is part of a recent eort to improve CTF capabilities for simulating flows in boiling nuclear reactors (BWRs). As it is of common knowledge among the thermodynamic community, the six-equation two-phase flow model is not strongly hyperbolic, meaning some of its eigenvalues are either not real or not distinct [7]. In a Cauchy sense and for an initial-value problem, strong hyperbolicity was found to be equivalent to well-posedness which ensures uniqueness of the solution in the space-time used. This concept is also of importance when designing high-order numerical methods and also when implementing boundary conditions that rely on real eigenvalues such as Riemann solvers. A tremendous amount of research is available in the literature on how to hyperbolize the six-equation two-phase flow model by adding ad hoc terms to the system of equations often referred to as virtual mass or interfacial pressure correction term. From a mathematical prospective, the main idea driving this approach is to add terms to the phasic momentum equation containing first-order partial derivatives of the unknowns to modify the Jacobian matrix and thus canceling the imaginary part of the eigenvalues.For this work it was chosen to implement a interfacial pressure correction (IPC) term on the same model as what was done for the CATHARE system code [8], that allows to recover a two-pressure formulation. In the following, the six-equation two-phase flow model is recalled along with the IPC term and its influence on the eigenvalues. Then, numerical results and convergence studies are presented for a water-faucet problem [9] at a pressure of 10 bar. Finally, conclusions and future work are addressed.

Delchini, Marco↗

NASA NTRSDate not supplied

Flow Boiling and Condensation Experiment Overview

A brief overview of the Flow Boiling and Condensation Experiment will be presented at the Technical Interchange Meeting between NASA-GRC Scientists and Engineers and the Korea Aerospace Research Institute (KARI). The presentation describes the development of the FBCE, its constraints, and the on-going testing that had been and continues to be executed onboard of the International Space Station (ISS).

International Space Station↗

NASA NTRS2016

Electrically Driven Liquid Film Boiling Experiment

This presentation presents the science background and ground based results that form the basis of the Electrically Driven Liquid Film Boiling Experiment. This is an ISS experiment that is manifested for 2021. Objective: Characterize the effects of gravity on the interaction of electric and flow fields in the presence of phase change specifically pertaining to: a) The effects of microgravity on the electrically generated two-phase flow. b) The effects of microgravity on electrically driven liquid film boiling (includes extreme heat fluxes). Electro-wetting of the boiling section will repel the bubbles away from the heated surface in microgravity environment. Relevance/Impact: Provides phenomenological foundation for the development of electric field based two-phase thermal management systems leveraging EHD, permitting optimization of heat transfer surface area to volume ratios as well as achievement of high heat transfer coefficients thus resulting in system mass and volume savings. EHD replaces buoyancy or flow driven bubble removal from heated surface. Development Approach: Conduct preliminary experiments in low gravity and ground-based facilities to refine technique and obtain preliminary data for model development. ISS environment required to characterize electro-wetting effect on nucleate boiling and CHF in the absence of gravity. Will operate in the FIR - designed for autonomous operation.

Electrohydrodynamics↗

NASA NTRS1988

Dynamics of face and annular seals with two-phase flow

Hughes, William F.↗

NASA NTRS2019

Flow Boiling and Condensation Experiment Flight Hardware Development

The Flow Boiling and Condensation Experiment (FBCE) to be manifested on the International Space Station (ISS) consists of a fluid system and the associated electronics to provide for conditioning the test fluid (normal-PerFluorohexane or nPFH-C6F14) to the proper thermodynamic state prior to entering a test module, which can be interchangeable based on the science objectives. Two separate test modules have been manufactured for the FBCE, the Flow Boiling Module (FBM), which investigates flow boiling for a subcooled liquid, saturated liquid, or two phase mixture, and the Condensation Module Heat Transfer (CM-HT), which investigates condensation of a flowing saturated or superheated vapor. The test fluid heating is accomplished using the Bulk Heater Module (BHM), which heats the fluid to various states based on the demands of the currently installed test module. ISS Internal Thermal Control System (ITCS) water is utilized to cool the test fluid prior to entering the circulation pump, and is also utilized for cooling for condensation in CM-HT, for cooling of a camera in FBM. An adjustable pressure bellows-type accumulator is used to set the pressure at the inlet of the test section, but does not provide active pressure control during testing. The flow of the test fluid is achieved using a gear pump controlled by a coriolis flow meter, which also provides the flow rate measurement. Flow rates for the ITCS water loops are measured and controlled using coriolis flow meters with directly controlled proportional valves. During execution of FBCE operations, the FBM is scheduled to collect data for three months before being exchanged with CM-HT for another three month data collection run. In this work, we present the development of the flight hardware, the associated challenges experienced during the development such as packaging flight system hardware, and the lessons learned in overcoming the encountered challenges.

Guzik, Monica C.↗

DOE OSTI2013

RELAP-7: Demonstrating the integration of two-phase flow components for an ideal BWR loop

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory. RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability (LWRS) Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). During Fiscal-Year 2013, a number of physical components have been developed to support the demonstration calculations presented in this report. The case selected for demonstration of an idealized BWR loop in RELAP-7 includes the major components for the primary system of a BWR. The BWR loop has been successfully simulated to steady state with the RELAP-7 code. The next major stage of development is to demonstrate two-phase flow modeling capability through a simplified boiling water reactor station blackout analysis, which will be reported in the subsequent demonstration simulation report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1974

Geothermal reservoir engineering research

The Stanford University research program on the study of stimulation and reservoir engineering of geothermal resources commenced as an interdisciplinary program in September, 1972. The broad objectives of this program have been: (1) the development of experimental and computational data to evaluate the optimum performance of fracture-stimulated geothermal reservoirs; (2) the development of a geothermal reservoir model to evaluate important thermophysical, hydrodynamic, and chemical parameters based on fluid-energy-volume balances as part of standard reservoir engineering practice; and (3) the construction of a laboratory model of an explosion-produced chimney to obtain experimental data on the processes of in-place boiling, moving flash fronts, and two-phase flow in porous and fractured hydrothermal reservoirs.

Ramey, H. J., Jr.↗

NASA NTRS2016

Fluid Studies on the International Space Station

Will discuss the recent activities on the international space station, including the adiabatic two phase flow, capillary flow and interfacial phenomena, and boiling and condensation. Will also give a historic introduction to Microgravity Studies at Glenn Research Center. Talk will be given to students and faculty at University of Louisville.

research↗

DOE OSTI2017

Laser-induced boiling of biological liquids in medical technologies

Using optical and acoustic methods we study thermal and transport processes related to the boiling of biological liquids under the action of continuous-wave laser radiation having moderate power (1 – 10 W) in the near-IR range (0.97 – 1.94 μm). These processes are investigated in the course of a few particular clinical procedures aimed at the modification and removal of pathological tissues (veins, mammary gland cyst, Baker’s cyst) and tissue regeneration (intervertebral discs). In the proposed approach, the modification and destruction of biotissues are due to the fast delivery of heat by two-phase jet flows, formed in the course of liquid boiling, rather than the direct laser heating. This provides the high rate of heat delivery to the pathological biotissue, avoiding its overheating (the temperature higher than 100 °C) and undesired heating of adjacent tissues. Two main regimes of laser-induced boiling near the optical fibre tip were revealed, namely, the heterogeneous jet boiling (arising when the fibre with a blackened tip is used) and the homogeneous boiling (with the radiation absorbed in the liquid volume). Both studied regimes allow one to obtain high specific heat flows, and the domination of one of the boiling regimes is determined by the presence of absorbing coating on the fibre tip, the tissue type, as well as by its shape (e.g., the presence of channels or cavities in the tissue). It is established that the heterogeneous jet boiling at the fibre tip corresponds to the regime of superintensive bubble boiling. (paper)

42 ENGINEERING↗

NASA NTRS1969

Test of sodium boiling in a single tube-in- shell heat exchanger over the range 1720 deg to 1980 deg F /1211 to 1355 K/

Sodium was boiled in a vertical, 1/2-inch (1.27-cm) inside diameter by 4-fOOt (1.22-m) long tube. The columbium alloy boiler operated at flow rates of 75 to 380 lbm/hr (9.4 to 48 g/sec) and up to 0.93 quality. Boiling performance depended on whether the inlet flow was two-phase or liquid. Two-phase momentum pressure drops were 1(1/2) times frictional drops. Qualities over 0.9 were obtained at some critical (burnout) boiling conditions. Liquid superheats up to 250° F (139 K) existed before initiation of boiling.

James P. Lewis↗

NASA NTRS1965

Alkali metals boiling and condensing investigations quarterly progress report no. 12, 1 apr. - 30 jun. 1965

Alkali metals boiling and condensing investigations of two-phase heat transfer and fluid flow for potassium

LIQUID FLOW↗

NASA NTRS1989

Dynamics of face and annular seals with two-phase flow

A detailed study was made of face and annular seals under conditions where boiling, i.e., phase change of the leaking fluid, occurs within the seal. Many seals operate in this mode because of flashing due to pressure drop and/or heat input from frictional heating. High pressure, water pumps, industrial chemical pumps, and cryogenic pumps are mentioned as a few of many applications. The initial motivation was the LOX-GOX seals for the space shuttle main engine, but the study was expanded to include any face or annular seal where boiling occurs. Some of the distinctive behavior characteristics of two-phase seals were discussed, particularly their axial stability. While two-phase seals probably exhibit instability to disturbances of other degrees of freedom such as wobble, etc., under certain conditions, such analyses are too complex to be treated at present. Since an all liquid seal (with parallel faces) has a neutral axial stiffness curve, and is stabilized axially by convergent coning, other degrees of freedom stability analyses are necessary. However, the axial stability behavior of the two-phase seal is always a consideration no matter how well the seal is aligned and regardless of the speed. Hence, axial stability is thought of as the primary design consideration for two-phase seals and indeed the stability behavior under sub-cooling variations probably overshadows other concerns. The main thrust was the dynamic analysis of axial motion of two-phase face seals, principally the determination of axial stiffness, and the steady behavior of two-phase annular seals. The main conclusions are that seals with two-phase flow may be unstable if improperly balanced. Detailed theoretical analyses of low (laminar) and high (turbulent) leakage seals are presented along with computer codes, parametric studies, and in particular a simplified PC based code that allows for rapid performance prediction. A simplified combined computer code for the performance prediction over the laminar and turbulent ranges of a two-phase seal is described and documented. The analyses, results, and computer codes are summarized.

Hughes, William F.↗

NASA NTRSDate not supplied

Mini-Channel Flow Boiling in 1g and μg

Subcooled flow boiling combines complex processes governed by a strong coupling of heat and mass transfer, hydrodynamics, and liquid-vapor transitions. The gravitational body force plays an important role in these processes. By exploiting the latent heat of vaporization, subcooled flow boiling can transfer up to several orders-of-magnitude more heat compared to forced liquid flow. It is widely used in Earth-based thermal equipment to reduce the cost, size, and power required to move heat. Subcooled flow boiling is now considered a promising technology for space applications since it offers a means to control the motion of bubbles near the heated surface and thereby enhance nucleate boiling. However, a lack of data and predictive models on the role that gravity plays in these phenomena hinders the use of flow boiling in a microgravity environment. We will present results of experiments on transients in nucleate boiling that were carried out in the Flow Boiling Module (FBM) on the International Space Station (ISS). The unique ISS FBM ability is to couple large heat loads with flow visualization in testing flow boiling under strict control of experimental conditions in long-duration microgravity. Comparison of results of microgravity experiments with data obtained in a similar setup on the NASA GRC demonstrates the role of gravity in nucleate boiling. The work is supported by the NASA Flight Opportunities program (grant 80NSSC21K0501) and NSF/CASIS program (NSF CBET grants 1832260, 2126461, 2126462).

Two-phase flow and phase transition↗

NASA NTRS2024

Microgravity Flow Boiling Experiments With Liquid-Vapor Mixture Inlet Onboard the International Space Station

This article is part of a series of studies culminating from the multi-objective Flow Boiling and Condensation Experiment (FBCE) onboard the International Space Station, which utilized the Flow Boiling Module (FBM) for experiments between February and July 2022. This study investigates microgravity flow boiling of n-Perfluorohexane with liquid-vapor mixture (two-phase) inlet conditions to the FBM with either one or two opposite walls heated. The FBM’s channel has a rectangular cross-sectional area of 5.0 × 2.5 mm 2 and a heated length of 114.6 mm. Key parameters of interest include mass velocity (180 – 2400 kg/m 2 s), inlet quality (-0.01 – 0.87), inlet pressure (120 – 200 kPa), and heat flux (1.8 W/cm 2 to critical heat flux), and a large database is amassed. The flow is visualized via a high-speed video camera and photographs are recorded at each heating increment to assess the periodic flow patterns within the channel and the near-wall interfacial behavior. Flow patterns are complex and mainly characterized by high- and low-density fronts alternately traversing the channel to yield high- and low-density-dominant periods of boiling. At all operating conditions, high-density fronts are faster during high-density-dominant periods. At low inlet qualities, the flow is annular near the channel inlet with a central vapor core surrounded by an annular liquid layer. Each high-density front having a high liquid fraction leaves a thin liquid layer sheared onto the heated walls. Boiling occurs within the liquid layer and a vapor layer is formed next to the heated wall. Inlet quality and mass velocity most dictate the overall flow patterns followed by heating configuration, and to a much lesser extent, heat flux and inlet pressure. Heat transfer characteristics are assessed via averaged boiling curves, streamwise profiles of wall temperature and heat transfer coefficient, and parametric curves of local and averaged heat transfer coefficients. Inlet pressure has an insignificant effect on heat transfer. At similar operating conditions, both the heating configurations yield similar trends and values of heat transfer coefficient and critical heat flux (CHF, slightly higher for single-sided) even though double-sided heating adds twice the heat to the fluid and doubly raises local quality. The heat fluxes required for both onset of nucleate boiling degradation and CHF are larger at high mass velocities and low inlet qualities. For a fixed inlet quality, high mass velocities yield higher average heat transfer coefficients at both lower and higher heat fluxes, while the nucleate boiling regime at intermediate heat fluxes is unaffected. For a fixed mass velocity, higher inlet qualities yield higher and lower average heat transfer coefficients at lower and higher heat fluxes, respectively.

Microgravity↗

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↗