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.

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.

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 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.

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.

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.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Evaluation of NDE Round-Robin Exercises Using the NRC Steam Generator Mockup at Argonne National Laboratory

2002 ◽  
Author(s):  
Joseph Muscara ◽  
David S. Kupperman ◽  
Sasan Bakhtiari ◽  
Jang-Yul Park ◽  
William J. Shack
Keyword(s):  
Steam Generator ◽  
Eddy Current ◽  
Fatigue Cracks ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Round Robin ◽  
Probability Of Detection ◽  
Maximum Depth ◽  
Tube Sheet ◽  
Intergranular Stress Corrosion

This paper discusses round-robin exercises using the NRC steam generator (SG) mock-up at Argonne National Laboratory to assess inspection reliability. The purpose of the round robins was to assess the current reliability of SG tubing inservice inspection, determine the probability of detection (POD) as function of flaw size or severity, and assess the capability for sizing of flaws. For the round robin and subsequent evaluation completed in 2001, eleven teams participated. Bobbin and rotating coil mock-up data collected by qualified industry personnel were evaluated. The mock-up contains hundreds of cracks and simulations of artifacts such as corrosion deposits and tube support plates that make detection and characterization of cracks more difficult in operating steam generators than in most laboratory situations. An expert Task Group from industry, Argonne National Laboratory, and the NRC have reviewed the signals from the laboratory-grown cracks used in the mock-up to ensure that they provide reasonable simulations of those obtained in the field. The mock-up contains 400 tube openings. Each tube contains nine 22.2-mm (7/8-in.) diameter, 30.5-cm (1-ft) long, Alloy 600 test sections. The flaws are located in the tube sheet near the roll transition zone (RTZ), in the tube support plate (TSP), and in the freespan. The flaws are primarily intergranular stress corrosion cracks (axial and circumferential, ID and OD) though intergranular attack (IGA) wear and fatigue cracks are also present, as well as cracks in dents. In addition to the simulated tube sheet and TSP the mock-up has simulated sludge and magnetite deposits. A multiparameter eddy current algorithm, validated for mock-up flaws, provided a detailed isometric plot for every flaw and was used to establish the reference state of defects in the mock-up. The detection results for the 11 teams were used to develop POD curves as a function of maximum depth, voltage and the parameter mp , for the various types of flaws. The POD curves were represented as linear logistic curves, and the curve parameters were determined by the method of Maximum Likelihood. The effect of both statistical uncertainties inherent in sampling from distributions and the uncertainties due to errors in the estimates of maximum depth and mp was investigated. The 95% one-sided confidence limits (OSL), which include errors in maximum depth estimates, are presented along with the POD curves. For the second round robin a reconfigured mock-up is being used to evaluate the effectiveness of eddy current array probes. The primary emphasis is on the X-Probe. Progress with the X-Probe round robin is discussed in this paper.

scholarly journals Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory

2002 ◽  
Author(s):  
Mary D. McDermott ◽  
Charles D. Griffin ◽  
Daniel K. Baird ◽  
Carl E. Baily ◽  
John A. Michelbacher ◽  
...  
Keyword(s):  
Sodium Hydroxide ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Test Facility ◽  
United States Department ◽  
Safe Condition ◽  
And Control ◽  
Low Level Radioactive Waste

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.

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