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Cryogenic fluid is not affected by zero gravity in the range of nucleate boiling but buoyant-force controlled processes of maximum heat flux, transition and film boiling are gravity sensitive
A model that is currently used to predict the priming and performance limitations of a monogroove heat pipe is expanded to include the boiling limitation and the cross-sectional temperature distribution as determined from a multidimensional finite element analysis technique. The improved model is verified experimentally and shown to accurately predict the cross-sectional temperature distribution when the heat flux distribution is known. The model provides a way to estimate the level at which nucleate boiling and the associated dryout of the capillary wick occurs.
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.
The reactivity-initiated accident (RIA) is a complex scenario with several tightly interacting physical phenomena. Accurately predicting fuel behavior during these transients is difficult due to limitations in the modeling of fuel-to-coolant heat transfer. Common approaches to simulate RIAs involve standalone calculations using either a fuel performance code or a thermal-hydraulic code. The complex interdependencies of thermal-hydraulic and fuel mechanical behavior suggest that a tight coupling between these codes may provide more accurate predictions of fuel-to-coolant heat transfer and cladding mechanical response. Here, RELAP5-3D and BISON are coupled in this paper to simulate RIAs, and a sensitivity analysis is performed to rank key thermal properties and two-phase heat transfer parameters relevant for fuel-to-coolant heat transfer and cladding failure mechanisms in UO 2 –Zircaloy-4 systems. Gas gap conductance, film boiling heat transfer uncertainty, pulse width, fuel-specific heat capacity, and cladding-specific heat capacity were identified as important parameters. Variations in figures of merit resulting from changes to pulse width and the material thermal properties indicate that time-dependent heat transfer rates are significant for safety-relevant mechanical parameters due to the time dependence of cladding ductility and pellet-cladding mechanical interaction loading. The results suggest that the thermal-hydraulic factors have a nonnegligible influence on the thermomechanical solution and vice versa. Tight coupling of both sets of physics is recommended to improve prediction of fuel behavior during RIAs. Highlights include the following: 1. The RELAP5-3D thermal-hydraulic code and the BISON fuel performance code are tightly coupled for simulation of RIA transients with energy depositions at the Zircaloy-4 cladding failure threshold. 2. Departure from nucleate boiling occurred for all simulated cases. Due to the ductility of fresh fuel, substantial ballooning occurred in most cases. 3. Gas gap conductance, fuel-specific heat capacity, cladding-specific heat capacity, transient pulse width, and film boiling heat transfer were the dominant thermal factors impacting the safety figures of merit at energy depositions.
The Flow Boiling and Condensation Experiment (FBCE) Transfer Line (TL) test module is being developed to study fluid physics of cryogenic transfer lines in microgravity. The TL module will reside in the FBCE facility located within the Fluids Integration Rack (FIR) aboard the International Space Station (ISS). TL will use the same normal perfluoro hexane (nPFH) used in the previous FBCE test sections. To simulate cryogenic line transfer in the visualization section, a sapphire tube coated with indium tin oxide (ITO) is heated via Joule heated while a bypass line is used to flow liquid nPFH at the desired inlet conditions. Once the visualization line reaches steady-state temperature and pressure conditions, flow is diverted through the heated line and chilldown begins. All flow regimes observed within a cryogenic TL are created and studied using visible imaging with diffuse backlighting to image inverted annular flow, transition, and nucleate boiling. A visible laser light sheet provided by FIR will provide a means to observe and track droplets in the dispersed flow regime. IR imaging of the visualization tube outer surface with the emissivity of the ITO coating provides the surface chilldown temperature profile. Preliminary imaging system development includes some of the components used in the Flow Boiling Module (FBM) and some new components that are easily integrated into the PC-based system but will require some refinement to operate properly on the FIR PCIe/104 imaging system. Optical design and packaging within the limited volume of the module enclosure is also presented. This paper presents FBCE-TL visible and infrared laboratory-based imaging studies and anticipated visualization capabilities of the TL module under preliminary breadboard development and under the utilization constraints within the FBCE/FIR host facility.
The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.
The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for thermal hydraulics analyses of research and test reactors. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to the fit functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one latera stripe of a fuel element. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.
The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software includes non-power research and test reactor analysis, and it has been used for the conversion to low- enriched uranium fuel of U.S. High Performance Research Reactors such as the Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat reactor safety analysis, such as during fuel reloading, STAT7 accommodates flexibility in analyzing many practical aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for reactor thermal hydraulics analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to fitted functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze every axial node in each lateral stripe of all fuel plates and coolant channels in every fuel element of an entire reactor core. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. In a single execution, STAT7 can be used to perform full core analysis by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.
With increasing per capita energy consumption and rapidly growing energy consumption predicted for data centers, the US faces challenges in meeting projected energy demands. The current administration is attempting to accelerate the deployment of new nuclear construction by providing significant investments in advanced nuclear and by modernizing regulatory approaches for the licensing new reactors. However, enabling extended power uprates (EPUs) for the current nuclear reactor fleet would provide a shorter-term solution to better calibrate near-term energy generation capacity to the ever-growing energy demands of the modern era. One pathway to achieve power uprates is to increase the operational window by reassessing fuel safety limits around anticipated transients. Existing light-water reactors (LWRs) utilize a targeted operational window with well-defined operational efficiencies, yet this window is also often bound by conservative safety limits that result in suboptimal operational performance. For example, the operation of the existing fleet is highly constrained by safeguards to operation related to anticipated operational occurrences (AOOs), which are moderate-frequency transients expected with a frequency greater than 0.01 per reactor-year. These AOOs include temperature excursions where the critical heat flux for the system is exceeded, resulting in a departure from nucleate boiling (DNB) at the cladding surface. This DNB event is a thermohydraulic condition that, with current conservative fuel safety limits, results in the fuel rods being classified as “failed,” meaning that the cladding may not be returned to service.
With increasing per capita energy consumption and rapidly growing energy consumption predicted for data centers, the US faces challenges in meeting projected energy demands. The current administration is attempting to accelerate the deployment of new nuclear construction by providing significant investments in advanced nuclear and by modernizing regulatory approaches for the licensing new reactors. However, enabling extended power uprates (EPUs) for the current nuclear reactor fleet would provide a shorter-term solution to better calibrate near-term energy generation capacity to the ever-growing energy demands of the modern era. One pathway to achieve power uprates is to increase the operational window by reassessing fuel safety limits around anticipated transients. Existing light-water reactors (LWRs) utilize a targeted operational window with well-defined operational efficiencies, yet this window is also often bound by conservative safety limits that result in suboptimal operational performance. For example, the operation of the existing fleet is highly constrained by safeguards to operation related to anticipated operational occurrences (AOOs), which are moderate-frequency transients expected with a frequency greater than 0.01 per reactor-year. These AOOs include temperature excursions where the critical heat flux for the system is exceeded, resulting in a departure from nucleate boiling (DNB) at the cladding surface. This DNB event is a thermohydraulic condition that, with current conservative fuel safety limits, results in the fuel rods being classified as “failed,” meaning that the cladding may not be returned to service.
Pool heating of liquid hydrogen in subcritical and supercritical pressure regimes over range of accelerations - effect on nucleate boiling
Heat transfer from cylindrical copper surface to liquid helium at 4 degrees K, discussing temperature fluctuations in nucleate boiling region
Liquid-vapor system in closed container, investigating transient pressure rise under heat and mass transfer interactions including incipient and nucleate boiling
The performance of an electrohydrodynamic heat pipe formed by enclosing three straight axial electrodes with a 1.27 cm O.D. internally threaded stainless steel tube was measured with Freon-11 as working fluid. The condenser and evaporator conductances at low power were comparable to those measured in a conventional capillary driven heat pipe using similar working fluids. The evaporator conductance improved at high power probably due to the onset of nucleate boiling. The maximum heat throughput for this pipe was substantially higher than that of a conventional pipe.
Inexpensive, safe apparatus built to test cooling by, ignition in, and nucleate boiling to film boiling of liquid oxygen flowing around ball bearings. Provides for heating of bearing specimens, direct observations, and measurements of temperatures and temperature distributions by thermocouples and infrared sensors. Used to evaluate suitability of various materials and surface treatments for ball bearings in high-pressure liquid-oxygen turbopumps. Ball rests in fixture while liquid oxygen flows around it. Flow reversed so observer or infrared sensor can view phenomena on trailing or leading side of ball.
A numerical model was developed, with the SINDA'85/FLUINT program, for calculating the thermal and hydrodynamic transients that occur during the chilldown of a cryogenic transfer line, using a well documented test case to validate the modeling process. Using this model, a total of ten cases were analyzed to evaluate the effects of variable inlet valve position, inlet pressures, and the use of an internal flow liner to promote nucleate boiling. It was found that an efficient transfer line cooldown can be achieved if the inlet flow is throttled, to reduce the flow rate and quality, and an internal flow liner such as Teflon is used.
The 'Tank Pressure Control Experiment/Thermal Phenomena (TPCE/TP)' is a reflight of the tank pressure control experiment (TPCE), flown on STS-43 in a standard Get-Away Special (GAS) container in August 1991. The TPCE obtained extensive video and digital data of the jet induced mixing process in a partially filled tank in low gravity environments. It also provided limited data on the thermal processes involved. The primary objective of the reflight of TPCE is to investigate experimentally the phenomena of liquid superheating and pool nucleate boiling at very low heat fluxes in a long duration low gravity environment. The findings of this experiment will be of direct relevance to space based subcritical cryogenic fluid system design and operation. Experiment hardware and results from the first TPCE are described in outline and graphic form.
Anodized blacks for suppressing stray light in optical systems can now be replaced by microscopically textured metal surfaces. An application of these black surfaces to the Clementine star-tracker navigational system, which will be launched in early 1994 to examine the Moon, en route to intercept an asteroid, is detailed. Rugged black surfaces with Lambertian BRDF less than 10(exp -2) srad(sup -1) are critical for suppressing stray light in the star-tracker optical train. Previously available materials spall under launch vibrations to contaminate mirrors and lenses. Microtextured aluminum is nearly as dark, but much less fragile. It is made by differential ion beam sputtering, which generates light-trapping pores and cones slightly smaller than the wavelength to be absorbed. This leaves a sturdy but light-absorbing surface that can survive challenging conditions without generating debris or contaminants. Both seeded ion beams and plasma immersion (from ECR plasmas) extraction can produce these microscopic textures without fragile interfaces. Process parameters control feature size, spacing, and optical effects (THR, BRDF). Both broad and narrow absorption bands can be engineered with tuning for specific wavelengths and applications. Examples are presented characterized by FTIR in reflection librators (0.95 normal emissivity), heat rejection, and enhanced nucleate boiling.