Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

2005 ◽  
Vol 152 (2) ◽  
pp. 162-169 ◽  
Author(s):  
Yong Hoon Jeong ◽  
Soon Heung Chang ◽  
Won-Pil Baek
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Vessel Wall ◽  
Reactor Vessel

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

2002 ◽  
Author(s):  
Ruwan K. Ratnayake ◽  
L. E. Hochreiter ◽  
K. N. Ivanov ◽  
J. M. Cimbala
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Visual Information ◽  
Two Phase Flow ◽  
Flow Behavior ◽  
Phase Flow ◽  
Spacer Grid ◽  
Two Phase ◽  
Spacer Grids

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained from operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids.

Experimental Study of Flow Critical Heat Flux in Low Concentration Water-Based Nanofluids

2008 ◽  
Author(s):  
Sung Joong Kim ◽  
Tom McKrell ◽  
Jacopo Buongiorno ◽  
Lin-Wen Hu
Keyword(s):  
Experimental Study ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Flow Boiling ◽  
Saturation Temperature ◽  
Outlet Temperature ◽  
Flow Loop ◽  
Maximum Heat ◽  
Low Concentration

Nanofluids are known as dispersions of nano-scale particles in solvents. Recent reviews of pool boiling experiments using nanofluids have shown that they have greatly enhanced critical heat flux (CHF). In many practical heat transfer applications, however, it is flow boiling that is of particular importance. Therefore, an experimental study was performed to verify whether or not a nanofluid can indeed enhance the CHF in the flow boiling condition. The nanofluid used in this work was a dispersion of aluminum oxide particles in water at very low concentration (≤0.1 v%). CHF was measured in a flow loop with a stainless steel grade 316 tubular test section of 5.54 mm inner diameter and 100 mm long. The test section was designed to provide a maximum heat flux of about 9.0 MW/m2, delivered by two direct current power supplies connected in parallel. More than 40 tests were conducted at three different mass fluxes of 1,500, 2,000, and 2,500 kg/m2sec while the fluid outlet temperature was limited not to exceed the saturation temperature at 0.1 MPa. The experimental results show that the CHF could be enhanced by as much as 45%. Additionally, surface inspection using Scanning Electron Microscopy reveals that the surface morphology of the test heater has been altered during the nanofluid boiling, which, in turn, provides valuable clues for explaining the CHF enhancement.

Critical Heat Flux in Heated Horizontal Thin Annuli

2000 ◽  
Author(s):  
R. M. Stoddard ◽  
M. F. Dowling ◽  
S. I. Abdel-Khalik ◽  
S. M. Ghiaasiaan ◽  
S. M. Jeter
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Test Facility ◽  
Inlet Temperature ◽  
Empirical Correction ◽  
Experimental Parameters ◽  
Vertical Test ◽  
Gap Width ◽  
High Equilibrium

Abstract The objectives of the work reported here were to experimentally study the critical heat flux in a heated thin, horizontal, annular flow passage cooled by subcooled water and to examine the applicability and relevance of the current predictive methods for critical heat flux to such passages. Experiments were performed in the Georgia Tech Microchannel Test Facility (GTMTF). The test section was an annulus with 6.45 and 7.77 mm inner and outer diameters, respectively (0.66 mm gap width), and an 18.5-cm long heated section. The experimental parameters investigated covered the following ranges: test section exit pressure: 0.344–1.034 MPa; coolant (water) mass flux: 100–380 kg/m2s; wall heat flux: 0.231–1.068 MW/m2; water inlet temperature: 30–65°C. The results, in agreement with the existing CHF data for large horizontal channels, indicated that CHF values were considerably lower than the expected CHF values for vertical test section configuration. In all the tests CHF occurred at relatively high equilibrium qualities, and was preceded by flow stratification which caused dryout of the upper surface of the flow channel. The data were correlated by introducing empirical correction multipliers into three widely-used correlations for vertical channels, and based on the compensated distortions method.

Critical Heat Flux Experiments and Correlation in a Long, Sodium-Heated Tube

1981 ◽  
Vol 103 (1) ◽  
pp. 74-80 ◽  
Author(s):  
D. M. France ◽  
R. D. Carlson ◽  
T. Chiang ◽  
W. J. Minkowycz
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Steam Generator ◽  
Test Section ◽  
Water Pressure ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Test Facility ◽  
Correlation Equations ◽  
Steam Generators

Critical heat flux (CHF) experiments were performed in the Steam Generator Test Facility (SGTF) at Argonne National Laboratory for application to liquid metal fast breeder reactor steam generators. The test section consisted of a single, straight, vertical, full-scale LMFBR steam generator tube with force-circulated water boiling upwards inside the tube heated by sodium flowing countercurrent in a surrounding annulus. The test section tube parameters were as follows: 10.1 mm i.d., 15.9 mm o.d., material = 2 1/4 Cr–1 Mo steel, and 13.1 m heated length. Experiments were performed in the water pressure range of 7.0 to 15.3 MPa and the water mass flux range of 720 to 3200 kg/m2˙s. The data exhibited two trends: heat flux independent and heat flux dependent. Empirical correlation equations were developed from over 400 CHF tests performed in the SGTF. The data and correlation equations were compared to the results of other CHF investigations.

Study on Critical Heat Flux in Narrow Rectangular Channel With Longitudinal Vortex

2010 ◽  
Author(s):  
Yan-Ping Huang ◽  
Jun Huang ◽  
Jian Ma ◽  
Qiu-Wang Wang
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Rectangular Channel ◽  
Longitudinal Vortex ◽  
Water Medium ◽  
Test Channel ◽  
High Heat Transfer ◽  
System Pressure

Longitudinal Vortex (LV) is produced by Longitudinal Vortex Generators (LVGs) with high heat transfer efficiency and acceptable pressure loss. Due to the relative long influence distance and simple structure, LVGs may be used in narrow channels with flat surface under high temperature and high pressure water medium, in this paper, the critical heat flux (CHF) is one of most important focus. The test channel has the size of 600 mm (length) × 40 mm (width) × 3 mm (height), was used to research the CHF characteristic of CHF affected by LVGs. The test channel is visual in three sides and remains one side for power supply. The LVGs used in the experiments are 14 mm (length) × 2.2 mm (width) × 1.8 mm (height) in dimensions, and periodically mounted on the inner wall of the steel plate. The parameters that are varied during the experiments as follows, system pressure from 0.43 to 0.85 MPa, inlet mass flow flux from 40.2 to 745.7 kg·m−2·s−1, inlet subcooling from 46.8 to 104.2 °C, exit quality from 0.183 to 0.997, surface heat flux from 0.294 to 2.316 MW·m−2. The experiments show that the CHF is improved by 24.3% while the total pressure drop through the test section is improved by 62.9%. The bubble growth and its evolutionary process in narrow rectangular channel with LVGs have been obtained during a short term when the CHF occurs, and it is found that the bubbles have been affected intensely by LV. Based on these experiment data, the growth and aggregation of bubbles have been depressed by LV, the mass, momentum and energy exchange between cold and hot areas in the test section have been strengthened. As a result, the heat transfer enhancement by LV can be explained by the destruction of thermal boundary layer.

Variation of Critical Heat Flux by Flow Oscillation in a Small Vertical Channel

2009 ◽  
Author(s):  
Toshihiro Murakami ◽  
Rei Takei ◽  
Tomio Okawa
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Flow Rate ◽  
Mass Flux ◽  
Test Section ◽  
Oscillation Period ◽  
Flow Oscillation ◽  
Oscillation Condition ◽  
Flux Condition ◽  
Heated Length

The effect of sinusoidal oscillation of inlet mass flux on the critical heat flux (CHF) in forced convective boiling was investigated in experiment and numerical calculation. In the experiment, the test section was a small stainless steel round tube of 5 mm in inside diameter, filtrated and deionized tap water was used as a test fluid, and the flow direction was set to vertical upward. The heated length was 1,600 mm. Electric power supplied to a circulation pump was varied periodically to oscillate the inlet mass flux sinusoidally. Direct current was passed through the test section tube to heat it ohmically. The occurrence of critical heat flux condition was detected using the signal from the thermocouples that were spot-welded on the outer wall of the test section tube. In the present experimental conditions, it was expected that the critical heat flux condition was triggered by the dryout of liquid film in annular two-phase flow regime. The main experimental parameters were the time-averaged inlet mass flux and the amplitude and period of flow oscillation. The system pressure was also used as an important experimental parameter since a boiling water reactor is operated under high pressure condition. If the oscillation period is long enough, it is expected that the critical heat flux under the flow oscillation condition is close to that for the steady state when the flow rate is equal to the minimum flow rate in the oscillatory condition. On the other hand, the decrease of the critical heat flux would be mitigated if the oscillation period is shortened, since interaction would take place between the thin and thick film regions within a boiling channel. In accordance with this expectation, the critical heat flux measured under the flow oscillation condition was reduced with an increase in the oscillation period. It was demonstrated that the reduction of critical heat flux under flow oscillation condition can be correlated fairly well using the concept of dimensionless heated length. Numerical calculations using a one-dimensional three-fluid model were also carried. The calculated critical heat fluxes for flow oscillation conditions increased with increased value of dimensionless heated length, as in the present experiment.

The Preliminary Experimental Study Based on Failure Mechanism of the RPV Lower Head

2016 ◽  
Author(s):  
XianKe Meng ◽  
LiKai Fei ◽  
Aijing Zhang ◽  
SiJiang Xiong ◽  
Lei Cui ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Failure Modes ◽  
Rectangular Channel ◽  
Model Development ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Containment Vessel ◽  
Lower Head

In-Vessel Retention is a key severe accident management strategy for reactors such as AP/CAP series reactors. The IVR success evaluation criterion is whether the RPV is melted through or not at the final RPV state. Once the RPV lower head melt through, the liquid corium will flow into the reactor cavity and will lead to complex phenomena, such us steam explosion and the reaction between the corium and concrete. These will make temperature and pressure of the containment vessel rise quickly and is a threat to the integrity of the containment vessel. When the wall surface of RPV lower head heating condition exceed the critical heat flux, the temperature rises rapidly, it is generally assumed that the RPV lower head in this state will inevitably melt through. This is the so-called IVR failure. In order to study the possible failure modes and mechanism of RPV lower head under the IVR measures, an experimental facility called TRECT is built. By measuring the parameters such as temperature, flow of the test section to study the influence to CHF by the parameters such as flow velocity and angle. All of these can provide reliable basis to the effectiveness appraisal and model development on the area of severe accident mitigation measures (IVR). To be specific, the test section is rectangular channel whose section is 50 × 20 mm. The upper surface is the heat surface and using a direct current heating mode to supply heat power. The heat flux can reach 1.5MW/m2. We use this upper surface heated rectangular channel to simulate RPV ERVC channel. By adjust the angle of test section to simulate the different circum ferential location of RPV lower head. And the Adjusting range can be 0° to 90°. The experimental results show that flow rate was reduced by 11% in the experiments, the critical heat flux density increased by 4.5%. Inclined angle increased from 16° to 29°, CHF increased by 7.9%.

Study on Saturated Flow Boiling Heat Transfer Under Vibration Conditions

2012 ◽  
Author(s):  
Yoshitaro Fujiyama ◽  
Hiroyasu Ohtake
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Low Voltage ◽  
Flow Boiling ◽  
Heating Surface ◽  
Nucleate Boiling ◽  
Circuit Board ◽  
Vertical Position ◽  
Saturated Flow

The ability to predict void formation, void fraction and critical heat flux —CHF— in flow boiling under oscillatory flow and vibration conditions is important to the safety technology of nuclear reactor during earthquake. In the present study, the onset of nucleate boiling —ONB— and CHF on saturated flow boiling under vibration conditions were investigated experimentally. Steady state experiments were conducted using a copper thin-film and saturated and subcooled water at 0.1 MPa. The liquid velocity was 0.25, 1.38, 3.20 and 4.07 m/s, respectively; the liquid subcooling was 0 K and 20 K. A heater was made of a printed circuit board. A test section was a rectangular flow channel of 10 mm width and 10 mm height. The test heater was heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The test section was arranged for horizontal position facing upward and for vertical position, respectively. For the vibration condition, the test section was set on a vibration table. The ONB was decided as an occurrence of the first boiling bubble. The critical heat flux was determined as that immediately before the heating surface physically burned-out. The CHF on saturated flow boiling under vibration conditions were investigated experimentally.

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