Gas–liquid two-phase flow in narrow horizontal annuli

Experimental data associated with the two-phase flow regimes, void fraction and pressure drop in horizontal, narrow, concentric annuli are presented. Two transparent test sections, one with inner and outer diameters of 6.6 and 8.6 mm, and an overall length of 46.0 cm; the other with 33.2 and 35.2 mm diameters and 43.0 cm length, respectively, were used. Near-atmospheric air and water constituted the gas and liquid phases, respectively. The gas and liquid superficial velocities were varied in the 0.02–57 and 0.1–6.1 m s⁻¹ ranges, respectively. The major two-phase flow patterns observed included bubbly, slug/plug, churn, stratified, and annular. Transitional regimes, where the characteristics of two distinct flow regimes could be observed in the test sections, included bubbly-plug, stratified-slug and annular-slug. The obtained flow regime maps were different than flow regime maps typical of large horizontal channels and microchannels with circular cross-sections. They were also different from the flow regimes in rectangular thin channels. The measured average void fractions for the two test sections were compared with predictions of several empirical correlations. Overall, a correlation proposed by Butterworth [Butterworth, D., 1975. A comparison of some void fraction relationships for co-current gas–liquid flow. Int. J. Multiphase Flow 1, 845–850] based on the results of Lockhart and Martinelli (1949) provided the most accurate prediction of the measured void fractions. The measured pressure drops were compared with predictions of several empirical correlations. The correlation of Friedel [Friedel, L., 1979. Improved friction pressure drop correlations for horizontal and vertical two-phase pipe flow. 3R Int. 18, 485–492] was found to provide the best overall agreement with the data.     

Countercurrent flow limitation in inclined channels with bends

The countercurrent flow limitation phenomenon in an inclined round channel connected to bends at both ends is experimentally studied. The channel inner diameter is 1.9 cm, the bend radius divided by channel diameter is 10, and the channel length divided by channel diameter is varied in the 105–315 range. Countercurrent flow rates are measured with liquid and gas superficial velocities in the 0.015–0.21 m s⁻¹ and 0.1–3.1 m s⁻¹ ranges respectively. Air and water at room temperature and atmospheric pressure are used in the experiments. The liquid injection system is a constant-head plenum, and the channel angle of inclination with respect to the vertical line is varied in the 0°–60° range. The measured liquid and gas flow rates for all the angles of inclination are correlated using Wallis' flooding correlation, with a unique value for each of the two constants in the correlation.     

Evolution of the structure of a gas–liquid two-phase flow in a large vertical pipe

The evolution of the structure of a gas–liquid flow in a large vertical pipe of 195 mm inner diameter was investigated at the TOPFLOW test facility in Rossendorf. Wire-mesh sensors were used to measure sequences of two-dimensional distributions of local instantaneous gas fraction within the complete pipe cross-section. The sensors own a resolution of 3 mm at a frequency of 2500 Hz. Superficial velocities were varied in a range covering flow regimes from bubbly to churn-turbulent flow. The distance between the gas injection and the sensor position was changed using a so-called variable gas injection system. It consists of six gas injection units, each equipped with three rings of injection orifices in the pipe wall (orifice diameter: 1 and 4 mm), which are fed from ring chambers. The gas flow towards these distributor chambers is individually controlled by valves. Measured bubble-size resolved radial gas fraction profiles reveal differences in the lateral migration of bubbles of different size starting from the injection at the wall. The evolution of bubble-size distributions allows to study bubble coalescence and break-up. The influence of the physical properties of the fluid was studied by comparing cold air–water experiments with steam–water tests at 65 bar.     

Interfacial measurements in horizontal stratified flow patterns

The interfacial characteristic parameters of horizontal stratified wavy flow patterns were experimentally investigated for a mixture of air and water two-phase flow by using the double-sensor, parallel wire conductance probe method. The experiments were conducted in a horizontal flow loop 15.4 m long consisting of Pyrex glass tubing of 50.3 mm i.d. The range of gas superficial velocities was from 0.85 to 31.67 ms⁻¹ and the liquid superficial velocities ranged from 0.014 to 0.127 ms⁻¹. Several interfacial wave patterns as described by Andritsos and Hanratty (Int. J. Multiphase Flow 13 (1987a) 583–603) were identified and their characteristic parameters such as wave height, most dominant frequency, mean propagation velocity and mean wavelength were investigated in terms of liquid and gas flow rates. The interfacial shear stress calculated from the experimental measurements was used to evaluate the most widely used interfacial shear models.     

An investigation of transient heat transfer in the region of flow boiling dryout with freon-12 in a heated tube

Power transient experiments using vertical round tube test sections have provided information on the heat transfer characteristics associated with a change from pre-dryout to post-dryout flow boiling conditions. The test sections were heated by passing electric current along the tube wall, and cooled internally by Freon-12 flowing upwards through the tube.Seven steel tubes of various sizes were used (internal diameters in the range 7.1–26.6 mm, wall thicknesses 0.9–2.0 mm, and lengths of 0.9–3.9 m). Data were obtained for coolant mass fluxes in the range 150–3270 kg m⁻² s⁻¹, at a nominal pressure of 1.0 MPa, with exit qualities in the range 0.3–1.0. The transients were initiated by small increases in power input to the test section. Heat transfer characteristics were determined by calculating wall temperature responses as functions of time and comparing these with the corresponding temperature traces recorded in the experiments.In relation to the temperature responses of the tube wall under these transient conditions, the results show that transition boiling has only a slight effect and that film boiling has very significant effects.     

Critical heat flux of water in vertical round tubes at low pressure and low flow conditions

An experimental study on critical heat flux (CHF) has been performed for water flow in vertical round tubes under low pressure and low flow (LPLF) conditions to provide a systematic data base and to investigate parametric trends. Totally 513 experimental data have been obtained with Inconel-625 tube test sections in the following conditions: diameter of 6, 8, 10 and 12 mm; heated length of 0.31.77 m; pressure of 106951 kPa; mass flux of 20277 kg m⁻² s⁻¹; and inlet subcooling of 50654 kJ kg⁻¹, thermodynamic equilibrium critical quality of 0.3231.251 and CHF of 1081598 kW m⁻². Flow regime analysis based on Mishima & Ishii’s flow regime map indicates that most of the CHF occurred due to liquid film dryout in annular-mist and annular flow regimes. Parametric trends are examined from two different points of view: fixed inlet conditions and fixed exit conditions. The parametric trends are generally consistent with previous understandings except for the complex effects of system pressure and tube diameter. Finally, several prediction models are assessed with the measured data; the typical mechanistic liquid film dryout model and empirical correlations of (Shah, M.M., 1987. Heat Fluid Flow 8 (4), 326–335; Baek, W.P., Kim, H.G., Chang, S.H., 1997. KAIST critical heat flux correlation for water flow in vertical round tubes, NUTHOS-5, Paper No. AA5) show good predictions. The measured CHF data are listed in Appendix B for future reference.     

Experimental investigation of local two-phase flow parameters of a subcooled boiling flow in an annulus

As a series of subcooling boiling flow tests, local two-phase flow parameters were obtained at SUBO (subcooled boiling) test facility under steam–water flow conditions. The test section is a vertical annulus of which the axial length is 4.165 m with a heater rod at the center of a channel. The inner and outer diameters of the test section and the heater rod are 35.5 mm and 9.98 mm, respectively. The test was performed by a two-stage approach. Stage-I for the measurement of local bubble parameters has been already done (Yun et al., 2009). The present work focused on the stage-II test for the measurement of local liquid parameters such as a local liquid velocity and a liquid temperature for a given flow condition of stage-I. A total of six test cases were chosen by following the test matrix of stage-I. The flow conditions are in the range of the heat flux of 370–563 kW/m², mass flux of 1110–2100 kg/(m² s) and inlet subcooling of 19–31 °C at pressure condition of 0.15–0.2 MPa. From the test, local liquid parameters were measured at 6 elevations along the test section and 11 radial locations of each elevation in addition to the previously obtained local void fraction, interfacial area concentration, Sauter mean diameter and bubble velocity. The present subcooled boiling (SUBO) data completes a data set for use as a benchmark, validation and model development of the Computational Fluid Dynamics (CFD) codes or existing safety analysis codes.     

Experimental data on burn-out in a simulated BWR fuel bundle

Critical heat flux (CHF) experiments have been carried out on a 16-rod test section having the typical geometry of boiling water reactor (BWR) fuel elements and in particular a 366 cm length. Heat fluxes were uniform, both axially and radially. The tests were carried out for the CNEN Plutonium Program on CISE's 8 MW IETI-3 facility, at 71 kg/cm² abs, mass velocities of 12–200 g/cm² s and inlet sub-cooling of 15–180°C. Each corner rod was instrumented with four separate thermocouples to detect nnd locate the initiation of CHF, while the other rods were instrumented with four-junction thermopiles.     

Sodium boiling experiments in a seven-pin bundle: Flow patterns and two-phase pressure drop

Two series of quasi-steady state sodium boiling experiments have been carried out in an electrically heated seven-pin bundle. The power levels (130–170 and 30–40 W/cm²) and other test conditions were selected to correspond to the core and radial breeder zones of a typical LMFBR. The test procedure involved the gradual reduction of mass flow rate through the bundle in a simulation of the consequences of a slowly growing blockage in the lower part of a reactor subassembly. By this means it was possible to study the development of quasi-steady state boiling up to the onset of permanent dryout. The results obtained provide information on flow regimes in the two-phase region, vapour qualities and flow rates at which cooling of the bundle can be effectively maintained, and the ultimate incidence of dryout. A relation for the two-phase pressure drop multiplier obtained from adiabatic pressure drop measurements in this geometry is given and compared with earlier results from single-channel geometry tests.     

TRACE simulation of sodium boiling in pin bundle experiments under loss-of-flow conditions

In the framework of PSI's FAST code system, the thermal–hydraulic code TRACE is being extended for representation of sodium two-phase flow. As the currently available version (v.5) is limited to the simulation of only single-phase sodium flow, its applicability range is not enough to study the behavior of a Generation IV sodium-cooled fast reactor (SFR) during transients in which boiling is anticipated. The work reported here concerns the extension of the non-homogeneous, non-equilibrium two-fluid models, which are available in TRACE for steam-water, to sodium two-phase flow simulation. The conventional correlations for ordinary gas–liquid flows are used as basis, with optional correlations specific to liquid metal where necessary. A number of new models for representation of the constitutive equations specific to sodium, with a particular emphasis on the interfacial transfer mechanisms, have been implemented and compared with the original closure models.A first assessment of the extended TRACE version has been carried out, by using the code to model experiments that simulate a loss-of-flow (LOF) accident in a SFR. One- and two-dimensional representations of the test section have been considered. Comparison of the 1D model predictions, with both experiment and SIMMER-III code predictions, confirm the ability of the extended TRACE code to predict the principal sodium boiling phenomena. Two-dimensional representation of the test section, however, has been found necessary for providing more detailed comparisons with the experimental data and thereby studying, in greater detail, the influence of the physical models on the calculated results.The paper thus presents a first-of-its-kind application of TRACE to two-phase sodium flow. It shows the capability of the extended code to predict sodium boiling onset, flow regimes, pressure evolution, dryout, etc. Although the numerical results are in good agreement with the experimental data, the physical models should be further improved. Other integral experiments are planned to be simulated, in order to further develop and validate the two-phase sodium flow modeling.     

Local measurement of interfacial area, interfacial velocity and liquid turbulence in two-phase flow

Double sensor probe and hotfilm anemometry methods were developed for measuring local flow characteristics in bubbly flow. The formulation for the interfacial area concentration measurement was obtained by improving the formulation derived by Kataoka and Ishii. The assumptions used in the derivation of the equation were verified experimentally. The interfacial area concentration measured by the double sensor probe agreed well with one by the photographic method. The filter to validate the hotfilm anemometry for measuring the liquid velocity and turbulent intensity in bubbly flow was developed based on removing the signal due to the passing bubbles. The local void fraction, interfacial area concentration, interfacial velocity, Sauter mean diameter, liquid velocity, and turbulent intensity of vertical upward air–water flow in a round tube with an inner diameter of 50.8 mm were measured by using these methods. A total of 54 data sets were acquired consisting of three superficial gas flow rates, 0.015–0.076 m s⁻¹, and three superficial liquid flow rates, 0.600, 1.00, and 1.30 m s⁻¹. The measurements were performed at the three locations: L/D=2, 32, and 62. This data is expected to be used for the development of reliable constitutive relations which reflect the true transfer mechanisms in two-phase flow.     

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m⁻¹ °C⁻¹ under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.     

FCI experiments in the corium/water system

The KROTOS fuel coolant interaction (FCI) tests are aimed at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a test series was performed using 3 kg of prototypical corium (80 w/o UO₂, 20 w/o ZrO₂) which was poured into a water column of ≤1.25 m in height (95 and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests were performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10–80 K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) were observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported alumina (Al₂O₃) test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing of the corium melt, an additional test was performed with a larger diameter test section. In all the corium tests an efficient quenching process (0.8–1.0 MW kg-melt⁻¹) with total fuel fragmentation (mass mean diameter 1.4–2.5 mm) was observed. Results from alumina tests under the same initial conditions are also given to highlight the differences in behaviour between corium and alumina melts during the melt/water mixing.     

Film boiling heat transfer from a vertical cylinder in forced flow of liquids under saturated and subcooled conditions at pressures

Forced convection film boiling heat transfer on a vertical 3-mm diameter and 180-mm length platinum test cylinder located in the center of the 40-mm inner diameter test channel was measured. Saturated water, and saturated and subcooled R113 were used as the test liquids that flowed upward along the cylinder in the test channel. Flow velocities ranged from 0 to 3 m s⁻¹, pressures from 102 to 490 kPa, and liquid subcoolings for R113 from 0 to 60 K. The heat transfer coefficients for a certain pressure and liquid subcooling are almost independent of flow velocity and of a vertical position on the cylinder for the flow velocities lower than ≈1 m s⁻¹ (the first range), and they become higher for the velocities higher than ≈1 m s⁻¹ (the second range). Slight dependence on a vertical position being nearly proportional to z^−1/4, where z is the height from the leading edge of the test cylinder, exists for the flow velocities in the second range. The heat transfer coefficients at each velocity in the first and second ranges are higher for higher pressure and liquid subcooling. Correlation for the forced convection film boiling heat transfer with radiation contribution on a vertical cylinder was derived by modifying an approximate analytical solution for a two-phase laminar boundary layer model to agree better with the experimental data. It was confirmed that the experimental data of film boiling heat transfer coefficients in water and R113 were described by the correlation within ±20% difference.     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.     

A study on the effect of gas flow rate on the wave characteristics in two-phase gas–liquid annular flow

Interfacial waves play a very important role in the mass, momentum and energy transport phenomena in annular flow. In this paper, film thickness time–trace measurements for air–water annular flow were collected in a small vertical tube using a parallel wire probe. Using the data, a typical disturbance wave shape was obtained and wave properties (e.g., width, height, speed and roughness) were presented. The liquid mass flux ranged from 100 to 200 kg/m² s and the gas mass flux ranged from 18 to 47 kg/m² s. Disturbance wave characteristics were defined and the effects of changing the gas flow rate on the wave spacing, wave width, wave peak height and wave base height were studied. An average velocity model for the wave and base regions has been developed to determine the wave velocity. The investigation method could be further extended to annular-mist flow which frequently occurs in boiling water reactors.     

Experimental studies on rewetting of hot vertical annular channel

Studies on the rewetting behaviour of hot vertical annular channels are of interest in the context of emergency core cooling in nuclear reactors following LOCA. Experimental studies were carried out to study the rewetting behaviour of a hot vertical annular channel, with hot inner tube, for bottom flooding and top flow rewetting conditions. The length of the inner tube of the test section was 3030 mm for bottom flooding rewetting experiments and 2630 mm for top flow rewetting experiments. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 lpm (11.7×10⁻⁵ m³ s⁻¹). The initial surface temperature of the inner tube was varied from 200 to 500°C. The experimental studies show that for a given initial surface temperature of the tube, the rewetting velocity increases with an increase in flow rate of water and it decreases with an increase in the initial surface temperature for a given water flow rate. For a given water flow rate and initial surface temperature, the rewetting velocity is higher in the case of rewetting under bottom flooding conditions as compared to that in the case of rewetting under top flow conditions. These conclusions agree with the conclusions reported in the earlier literature. Using the experimental data of the present work, correlations for bottom flooding and top flow rewetting velocities are developed.     

Novel experimental measuring techniques required to provide data for CFD validation

CFD code validation requires experimental data that characterize the distributions of parameters within large flow domains. On the other hand, the development of geometry-independent closure relations for CFD codes have to rely on instrumentation and experimental techniques appropriate for the phenomena that are to be modelled, which usually requires high spatial and time resolution. The paper reports about the use of wire-mesh sensors to study turbulent mixing processes in single-phase flow as well as to characterize the dynamics of the gas–liquid interface in a vertical pipe flow. Experiments at a pipe of a nominal diameter of 200 mm are taken as the basis for the development and test of closure relations describing bubble coalescence and break-up, interfacial momentum transfer and turbulence modulation for a multi-bubble-class model. This is done by measuring the evolution of the flow structure along the pipe. The transferability of the extended CFD code to more complicated 3D flow situations is assessed against measured data from tests involving two-phase flow around an asymmetric obstacle placed in a vertical pipe. The obstacle, a half-moon-shaped diaphragm, is movable in the direction of the pipe axis; this allows the 3D gas fraction field to be recorded without changing the sensor position. In the outlook, the pressure chamber of TOPFLOW is presented, which will be used as the containment for a test facility, in which experiments can be conducted in pressure equilibrium with the inner atmosphere of the tank. In this way, flow structures can be observed by optical means through large-scale windows even at pressures of up to 5 MPa. The so-called “Diving Chamber” technology will be used for Pressurized Thermal Shock (PTS) tests. Finally, some important trends in instrumentation for multi-phase flows will be given. This includes the state-of-art of X-ray and gamma tomography, new multi-component wire-mesh sensors, and a discussion of the potential of other non-intrusive techniques, such as neutron radiography and magnetic resonance imaging (MRI).     

Experimental Results for an Acoustic Driver for MTF

General Fusion is planning to form an FRC or spheromak of 10¹⁷ cm⁻³, 100 eV, 40 cm diameter by merging two spheromaks with reverse or co-helicity. This target will be further compressed in a 3 m diameter tank filled with liquid PbLi with the plasma in the center. The tank is surrounded with pneumatically powered impact pistons that will send a convergent shock wave in the liquid to compress the plasma to 10²⁰ cm⁻³, 10 keV, 4 cm diameter for 7 μs. General Fusion has built a 500 kJ, 80 μs, 6 GW pneumatic impact piston capable of developing 2 GPa (300 kpsi). In this paper we will present the performances achieved to date.     

Corrosion resistance of stainless chrome-nickel steels in sodium as a function of its oxygen content

The article describes an experimental apparatus for studying the corrosion resistance of construction materials in circulating liquid sodium, as well as methods for the continuous removal of oxides from the sodium by means of porous metal filters and for the determination of its oxygen content. It describes the results of corrosion experiments and measurements of mechanical properties performed on six brands of stainless steel, specimens of which were placed in sodium containing 3·10^–3% and (4–5)·10^–2% oxygen at a temperature of 550°. The flow rate was 1.5 m/sec. The results of the corrosion experiments indicate that the corrosion rate of Type 18CrSNi stainless steels in sodium containing as much as (4–5)·10^–2% oxygen does not exceed 21.3 mg/dm² per month; these steets are not subject to intergranular corrosion, and their mechanical properties remain virtually machanged.G. V. Akimov State Scientific Research Institute for the Preservation of Materials, Prague, Czechoslovak Soviet Socialist Republic Translated from Atomnaya Énergiya, Vol. 14, No. 4, pp. 375–382, April, 1963