Nucleate pool boiling of saturated Freon 113 in a reduced gravity environment
Nucleate pool boiling of saturated Freon 113 in reduced gravity environment
Littles, J. W.↗
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Combined roles of buoyancy and orientation in nucleate pool boiling
The paper presents results obtained during pool boiling from flat surfaces over body force levels of a/g = -1 to 0 to 20 for LN2 and R-113 under steady and transient conditions. Pyrex and metal surfaces are used. Particular attention is given to the case where the effective gravity is perpendicular to the heating surface. At the lower levels of heat flux with nucleate boiling, the heater surface superheat first increases and then decreases as the body force increases above a/g = 1. This is due to the greater contribution of nonboiling convection.
Merte, H., Jr.↗
Critical heat flux on zircaloy and accident tolerant fuel cladding under prototypical conditions of pressurized and boiling water reactors
First of a kind high spatial and temporal resolution temperature measurements of Pressurized Water Reactor and Boiling Water Reactor simulated fuel pins (high heat flux cosine profile heaters) in annular rod type geometry during nucleate boiling, departure from nucleate boiling and re-wetting has been performed. Distributed fiber optic measurements of the cladding axial temperature with spatial resolutions down to 2.5 mm axially at up to three equally spaced azimuthal locations per rod have demonstrated the similarities and differences of the different accident tolerant fuel cladding materials. These experiments show the performance of accident tolerant cladding with a precise measurement of the location and evolution of dryout followed by the re-wetting phenomena (see supplementary time history of a critical heat flux event). Experimental parameters varied coolant mass fluxes from 1695 to 2712 kg/m 2 s, pressures from 10 to 20 MPa, and inlet subcooling from 10 to 55 °C. It was found that the coatings and different material have little effect on the value of the critical heat flux. Any discrepancies between the local critical heat flux verified among the cladding material are attributed to different departure from nucleate boiling phenomena captured during the experiment with the high-resolution optical fiber temperature data. Substantial differences in the performance of the different cladding material to survive a critical heat flux event were however observed by post-test X-ray and optical inspection. External evident damage and internal cracks to the simulated fuel pin having the bare zircaloy cladding were observed in regions associated with the occurrence of the critical heat flux for longer periods. Such damages were associated with zirconium oxidation, and not noticed on the Cr coated zircaloy or FeCrAl claddings. These results indicate that the use of FeCrAl or a Cr coated zircaloy as cladding in Pressurized Water Reactor and Boiling Water Reactor may increase the survivability of the cladding material for short transients.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
Boiling from small cylinders.
Heat transfer is observed as a function of temperature on small horizontal wires in water and four organic liquids. When the wire radius is sufficiently small, the hydrodynamic transitions in the boiling curve disappear and the curve becomes monotonic. Three modes of heat removal are identified for the monotonic curve and described analytically: a natural convection mode, a mixed film boiling and natural convection mode, and a pure film boiling mode. Nucleate boiling does not occur on the small wires. The study was motivated by an interest in predicting the behavior of large heaters at low gravity. The application of the present results to such circumstances is therefore discussed. It is proposed that the peak and minimum heat fluxes will vanish at low gravity as well as on small wires.
Bakhru, N.↗
A User’s Guide to the PLTEMP/ANL Code
PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A User’s Guide to the PLTEMP/ANL Code (V.4.3)
PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It should be noted that for a 3-dimensional solution the axial power shape is modeled without accounting for axial heat conduction. It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A User’s Guide to the PLTEMP/ANL Code
PLTEMP/ANL V4.4 is a program that obtains a steady-state flow and temperature solution, in geometrical extent, for a nuclear reactor core, or a single fuel assembly, including the effect of manufacturing and modelling uncertainties by the means of hot channel factors. It is based on an evolutionary sequence of codes originally used for calculating plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over the past 30 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability for select design and safety parameters is available to greatly reduce the reactor analyst’s time.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Numerical Simulation of Liquid Nitrogen Chilldown of a Vertical Tube
This paper presents the results of a one-dimensional numerical simulation of the transient chilldown of a vertical stainless steel tube with liquid nitrogen. The direction of flow is downward (with gravity) through the tube. Heat transfer correlations for film, transition, and nucleate boiling, as well as critical heat flux, rewetting temperature, and the temperature at the onset of nucleate boiling were used to model the convection to the tube wall. Chilldown curves from the simulations were compared with data from 55 recent liquid nitrogen chilldown experiments. With these new correlations the simulation is able to predict the time to rewetting temperature and time to onset of nucleate boiling to within 25% for mass fluxes ranging from 61.2 to 1150 kg/(sq m s), inlet pressures from 175 to 817 kPa, and subcooled inlet temperatures from 0 to 14 K below the saturation temperature.
GFSSP↗
Enhanced Boiling on Micro-Configured Composite Surfaces Under Microgravity Conditions
In order to accommodate the growing thermal management needs of future space platforms, several two-phase active thermal control systems (ATCSs) have evolved and were included in the designs of space stations. Compared to the pumped single-phase liquid loops used in the conventional Space Transportation System and Spacelab, ATCSs offer significant benefits that may be realized by adopting a two-phase fluid-loop system. Alternately, dynamic power systems (DPSs), based on the Rankine cycle, seem inevitably to be required to supply the electrical power requirements of expanding space activities. Boiling heat transfer is one of the key technologies for both ATCSs and DPSs. Nucleate boiling near critical heat flux (CHF) can transport very large thermal loads with much smaller device size and much lower pumping power. However, boiling performance deteriorates in a reduced gravity environment and operation in the CHF regime is precarious because any slight overload will cause the heat transfer to suddenly move to the film boiling regime, which in turn, will result in burnout of the heat transfer surfaces. New materials, such as micro-configured metal-graphite composites, can provide a solution for boiling enhancement. It has been shown experimentally that this type of material manifests outstanding boiling heat transfer performance and their CHF is also extended to higher values. Due to the high thermal conductivity of graphite fiber (up to 1,200 W/m-K in the fiber direction), the composite surfaces are non-isothermal during the boiling process. The composite surfaces are believed to have a much wider safe operating region (a more uniform boiling curve in the CHF regime) because non-isothermal surfaces have been found to be less sensitive to variations of wall superheat in the CHF regime. The thermocapillary forces formed by the temperature difference between the fiber tips and the metal matrix play a more important role than the buoyancy in the bubble detachment, for the bubble detachment manifests itself by a necking process which should not be weakened by reduced gravity. In addition, the composite surfaces introduce no extra pressure drop, no fouling and do not impose significant primary or maintenance costs. All of these suggest that this type of composite is an ideal material for the challenge of accounting for both reliability and economy of the relevant components applied in the ATCSs, the DPSs and other devices in future space missions. The aim of the proposed work is to experimentally investigate high nucleate pool boiling performance on a micro-configured metal-graphite composite surface and to determine the mechanisms of the nucleate boiling heat transfer both experimentally and theoretically. Freon-113 and water will be used as the test liquids to investigate wettability effects on boiling characteristics. The Cu-Gr and Al-Gr composites with various volume fractions of graphite fibers will be tested to obtain the heat transfer characteristic data in the nucleate boiling region and in the CHF regime. In the experiments, the bubble emission and coalescence processes will be recorded by a video camera with a magnifying borescope probe immersed in the working fluid. The temperature profile in the thermal boundary layer on the composite surfaces will be measured by a group of micro thermocouples consisting of four ultra fine micro thermocouples. This instrument was developed and successfully used to measure the temperature profile of evaporating liquid thin layers by the proposers in a study performed at the NASA/Lewis Research Center. A two tier model to explain the nucleate boiling process and the performance enhancement on the composite surfaces has been suggested by the authors. According to the model, the thicknesses of the microlayer and the macrolayer underneath the bubbles and mushrooms, can be estimated by the geometry of the composite surface. The experimental results will be compared to the predictions from the model, and in turn, to revise and improve it.
Zhang, Nengli↗
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↗
Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor
This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.
Nguyen, Kien-Cuong↗
The STAT7 Code for Statistical Propagation of Uncertainties in Steady State Thermal Hydraulics Analysis of Plate Fueled Reactors
The STAT7 software was developed to perform steady-state thermal hydraulic analyses. 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 MITR-II. 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 common fabrication parameters and other key reactor analysis uncertainties required for research and test reactor thermal hydraulic analyses. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one lateral stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. 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 element in the core.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Effects of surface conditions on nucleate pool boiling of sodium.
Commercial grade sodium boiling from horizontal disk at various pressures and temperatures attaining high heat fluxes, noting surface finish effects
NUCLEATE BOILING↗
Transition from Pool to Flow Boiling: The Effect of Reduced Gravity
Applications of boiling heat transfer in space can be found in the areas of thermal management, fluid handling and control, power systems, on-orbit storage and supply systems for cryogenic propellants and life support fluids, and for cooling of electronic packages for power systems associated with various instrumentation and control systems. Recent interest in exploration of Mars and other planets, and the concepts of in-situ resource utiliLation on Mars highlights the need to understand the effect of gravity on boiling heat transfer at gravity levels varying from 1>= g/g(sub e) >=10(exp -6). The objective of the proposed work was to develop a mechanistic understanding of nucleate boiling and critical heat flux under low and micro-gravity conditions when the velocity of the imposed flow is small. For pool boiling, the effect of reduced gravity is to stretch both the length scale as well as the time scale for the boiling process. At high flow velocities, the inertia of the liquid determines the time and the length scales and as such the gravitational acceleration plays little role. However, at low velocities and at low gravity levels both liquid inertia and buoyancy are of equal importance. At present, we have little understanding of the interacting roles of gravity and liquid inertia on the nucleate boiling process. Little data that has been reported in the literature does not have much practical value in that it can not serve as a basis for design of heat exchange components to be used in space. Both experimental and complete numerical simulations of the low velocity, low-gravity nucleate boiling process were carried out. A building block type of approach was used in that first the growth and detachment process of a single bubble and flow and heat transfer associated with the sliding motion of the bubble over the heater surface after detachment was studied. Liquid subcooling and flow velocity were varied parametrically. The experiments were conducted at 1 g(sub e), while varying the orientation of surface with respect to the gravity vector. In the laboratory experiments, holographic interferometry was used to obtain data on velocity and temperature fields associated with a bubble prior to, and after detachment and during sliding motion. A test rig for conducting experiments in the KC-135 was developed, but experiments could not be conducted due to the unavailability of the aircraft prior to completion of the project. Numerical simulations modeling the micro and macro regions of the bubble were carried out in three dimensions. The results of the experiments were used to validate analytical/numerical models.
Dhir, Vijay K.↗
Effects of surface conditions on nucleate pool boiling of sodium.
Commercial grade sodium boiling from horizontal disk at various pressures and temperatures attaining high heat fluxes, noting surface finish effects
SODIUM↗
Elucidating the Impact of Flow on Material-Sensitive Critical Heat Flux and Boiling Heat Transfer Coefficients: An Experimental Study with Various Materials
In this work, steady-state internal flow boiling critical heat flux (CHF) experiments were carried out at 10°C inlet subcooling and atmospheric pressure (84 kPa at an altitude of about 5300 feet) with mass flux ranging from 200 to 2,000 kg/m 2 -s on seven tube materials, including the accident-tolerant fuel cladding candidate FeCrAl (C26M) alloy. At a relatively low flow rate, the sensitivity of CHF to different materials was appreciable (the average difference among the tested materials was 19% at 200 kg/m 2 -s). As the absolute CHF differences stayed constant for higher flow rates, the increased flow rate led to reduced relative CHF differences among the tested materials (average relative difference of 15% at 1,000 kg/m 2 -s). FeCrAl alloy claddings that were oxidized under simulated Light Water Reactor (LWR) conditions yielded CHF comparable to as-received FeCrAl alloy cladding, with less than 1.8% difference. Experimental CHF results imply marginal material sensitivity impacts on the departure from nucleate boiling ratio in steady-state nuclear reactor cores. Yet, for accident progression analysis, material sensitivity may hold importance, as boiling behavior variation at low mass flux is appreciable. Macroscopic material parameters (surface wettability, surface roughness, and thermal properties) do not provide a suitable explanation for observed material-sensitive CHFs. Yet, surface roughness and thermal effusivity revealed power and linear relations, respectively, to the nucleate boiling heat transfer coefficients at low mass flux. At high mass flux, material sensitivity on nucleate boiling heat transfer significantly decreased due to the increased shear stress on boiling surface overwhelming material-related boiling implications.
36 MATERIALS SCIENCE↗
Transient boiling heat transfer in saturated liquid nitrogen and F113 at standard and zero gravity
Transient and steady state nucleate boiling in saturated LN2 and F113 at standard and near zero gravity conditions were investigated for the horizontal up, vertical and horizontal down orientations of the heating surface. Two distinct regimes of heat transfer mechanisms were observed during the interval from the step increase of power input to the onset of nucleate boiling: the conduction and convection dominated regimes. The time duration in each regime was considerably shorter with LN2 than with F113, and decreased as heat flux increased, as gravity was reduced, and as the orientation was changed from horizontal up to horizontal down. In transient boiling, boiling initiates at a single point following the step increase in power, and then spreads over the surface. The delay time for the inception of boiling at the first site, and the velocity of spread of boiling varies depending upon the heat flux, orientation, body force, surface roughness and liquid properties, and are a consequence of changes in boundary layer temperature levels associated with changes in natural convection. Following the step increase in power input, surface temperature overshoot and undershoot occur before the steady state boiling temperature level is established.
Oker, E.↗
Zero-G Report ''LH2 Boiling Threshold"
Test to determine minimum temperature difference between heated wall and liquid hydrogen at boiling point required to initiate nucleate boiling under steady state conditions and zero-g.
WALL TEMPERATURE↗