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.

Development of an Experimental Facility for Investigating Critical Heat Flux of Saturated Flow Boiling of Refrigerant-123 in Microchannels

2005 ◽  
Author(s):  
Wai Keat Kuan ◽  
Satish G. Kandlikar
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Flow Boiling ◽  
Test Facility ◽  
Uniform Heat Flux ◽  
Inlet Temperature ◽  
Experimental Facility ◽  
Copper Block ◽  
Saturated Flow ◽  
Cartridge Heater

An experimental facility is developed to investigate critical heat flux (CHF) of saturated flow boiling of Refrigerant-123 (R-123) in microchannels. Six parallel Microchannels with cross sectional area of 0.2 mm × 0.2 mm are fabricated on a copper block, and a Polyvinyl Chloride (PVC) cover is then placed on top of the copper block to serve as a transparent cover through which flow patterns and boiling phenomena could be observed. A resistive cartridge heater is used to provide a uniform heat flux to the microchannels. The experimental test facility is designed to accommodate test sections with different microchannel geometries. The mass flow rate, inlet pressure, inlet temperature of Refrigerant-123, and the electric current supplied to the resistive cartridge heater are controlled to provide quantitative information near the CHF condition in microchannels. A high-speed camera is used to observe and interpret flow characteristics of CHF condition in microchannels.

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.

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 During Natural Circulation Boiling of Saturated Liquid in Annulus With Uniformly Heated Outer Tube

1999 ◽  
Vol 122 (1) ◽  
pp. 74-79 ◽  
Author(s):  
M. Monde ◽  
Y. Mitsutake
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Natural Circulation ◽  
Outer Tube ◽  
Tube Length ◽  
Equivalent Diameter ◽  
Heated Tube ◽  
Annular Gap ◽  
Gap Width ◽  
Annular Tube

Critical heat flux has been measured during natural circulation boiling of water and R113 on a uniformly heated outer tube in a vertical annular tube. The experiment was carried out using water at atmospheric pressure and R113 at a pressure of 0.1–0.4 MPa for the annular gap width of S=1.0-4.0 mm, the heated tube diameter of 9–17 mm, and the annular tube length of 100–1000 mm. The similarity of critical heat flux between annular configurations of either inner or outer heated tubes and a simple heated tube can be clearly elucidated based on the characteristics of the heated equivalent diameter. The critical heat flux measured for S=1 mm can be predicted accurately by existing correlation for the annular tube and for clearance larger than S=4 mm by existing correlation for the single tube. A new correlation for medium clearances from S=1 to 4 mm has been developed to connect between both the existing correlations. [S0022-1481(00)01901-0]

Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis

2020 ◽  
Author(s):  
Robert Armstrong ◽  
Charles Folsom ◽  
Connie Hill ◽  
Colby Jensen
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Test Facility ◽  
Inverse Heat Transfer ◽  
Transfer Method ◽  
Nuclear Reactor Safety ◽  
Accident Scenarios

Abstract Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test facility (TREAT) at Idaho National Laboratory (INL) was developed. The experiment, named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless-steel heater rod submerged in a static water pool heated via the (n, α) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software Dakota to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.

Investigation of Critical Heat Flux in Single Fuel Pin With and Without Wire Spacer

2014 ◽  
Author(s):  
Tri Dan Le ◽  
Noriaki Inaba ◽  
Minoru Takahashi
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Inlet Temperature ◽  
Small Range ◽  
Two Phase ◽  
Mass Fluxes ◽  
Fuel Pin ◽  
High Conversion Ratio ◽  
Phase Water ◽  
Tight Lattice

Light water reactor could have fast neutron spectrum with high conversion ratio nearly equal unity by using tight lattice fuel assembly with wire spacer. There were some previous about critical heat flux for tight lattice but it were not focused on small range of qualities and also not directly using water as a coolant. We experimentally simulated vertical single fuel rod geometry with and without a wire spacer by using an electrically heated stainless steel rod. The rod is cooled by single or two-phase water in vertical up flow (from the bottom to the top) depending on the electrical input. We determined the critical heat flux for this system by varying the inlet temperature from 333 to 373 K and mass fluxes from 205 to 410 kg/m2s. The result show the critical heat flux (CHF) data in two phase flow condition base on inlet and outlet condition in both case of heater pin with and without wire. The CHF values were higher with wire than without wire due to the effect of wire and spiral flow.

AREVA’s Test Facility KATHY: Robust Critical Heat Flux Measurements, a Prerequisite for Reliable CHF Prediction

2014 ◽  
Author(s):  
O. Wieckhorst ◽  
J. Kronenberg ◽  
H. Gabriel ◽  
S. Opel ◽  
D. Kreuter ◽  
...  
Keyword(s):  
Heat Flux ◽  
Fuel Assembly ◽  
Critical Heat Flux ◽  
Development Process ◽  
Heat Removal ◽  
Test Facility ◽  
Hydraulic Facility ◽  
Key Factor ◽  
Unique Experience ◽  
High Level

The primary tool for assuring the heat removal from the fuel design’s rod surfaces is properly represented in the numerical simulations of a LWR fuel assembly design is the critical heat flux (CHF) or dryout correlation. During the last decade, AREVA has compiled unique experience in correlation development that has led to an improved development process to meet increased technical challenges. This is based upon the high level of expertise in CHF measurements for PWR and BWR fuel assembly designs gained by AREVA at its KATHY facility (KArlstein Thermal HYdraulic facility). The utilization of KATHY in conjunction with this improved development process is a key factor in ensuring reliable CHF prediction for safety analysis application. This paper describes the capabilities of the KATHY loop and the process used by AREVA to attain high quality CHF measurements.

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