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Critical Heat Flux for Downward-Facing Boiling on a Coated Hemispherical Vessel Surrounded by an Insulation Structure

Description: An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This nonmonotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.
Date: May 1, 2005
Creator: Yang, J.; Cheung, F. B.; Rempe, J. L.; Suh, K. Y. & Kim, S. B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Final Report: Investigation of Boiling Flow Regimes and Critical Heat Flux

Description: From abstract: A program to investigate the mechanism of the critical heat flux condition from the standpoint of flow regimes has been initiated at Dynatech for the AEC. This report covers the work done on this investigation in the first year.
Date: March 1, 1965
Creator: Suo, M.; Bergles, Arthur E.; Doyle, Edward F.; Clawson, L. & Goldberg, P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Critical Heat Flux in Inclined Rectangular Narrow Gaps

Description: In light of the TMI-2 accident, in which the reactor vessel lower head survived the attack by molten core material, the in-vessel retention strategy was suggested to benefit from cooling the debris through a gap between the lower head and the core material. The GAMMA 1D (Gap Apparatus Mitigating Melt Attack One Dimensional) tests were conducted to investigate the critical heat flux (CHF) in narrow gaps with varying surface orientations. The CHF in an inclined gap, especially in case of the downward-facing narrow gap, is dictated by bubble behavior because the departing bubbles are squeezed. The orientation angle affects the bubble layer and escape of the bubbles from the narrow gap. The test parameters include gap sizes of 1, 2, 5 and 10 mm and the open periphery, and the orientation angles range from the fully downward-facing (180o) to the vertical (90o) position. The 15 ×35 mm copper test section was electrically heated by the thin film resistor on the back. The heater assembly was installed to the tip of the rotating arm in the heated water pool at the atmospheric pressure. The bubble behavior was photographed utilizing a high-speed camera through the Pyrex glass spacer. It was observed that the CHF decreased as the surface inclination angle increased and as the gap size decreased in most of the cases. However, the opposing results were obtained at certain surface orientations and gap sizes. Transition angles, at which the CHF changed in a rapid slope, were also detected, which is consistent with the existing literature. A semi-empirical CHF correlation was developed for the inclined narrow rectangular channels through dimensional analysis. The correlation provides with best-estimate CHF values for realistically assessing the thermal margin to failure of the lower head during a severe accident involving relocation of the core material.
Date: June 1, 2004
Creator: Kim, Jeong J.; Kim, Yong H.; Kim, Seong J.; Noh, Sang W.; Suh, Kune Y.; Rempe, Joy L. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis.

Description: Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF occurs as a function of channel flow rate. The two sets of functions are combined to yield predictions of the power at which CHF occurs in the reactor. A combination of the RELAP5-3D code and the 2006 Groeneveld table predicts 67% more CHF power than does a combination of the STAT code and the Bernath correlation. Replacing the 2006 Groeneveld table with the Bernath CHF correlation (while using the RELAP5-3D code flow solution) causes the increase to be 23% instead of 67%. Additional RELAP5-3D flow-versus-power solutions obtained from Reference 1 and presented in Appendix B for four specific TRIGA reactors further demonstrates that the Bernath correlation predicts CHF to occur at considerably lower power levels than does the 2006 Groeneveld table. Because of the lack of measured CHF data in the region of interest to TRIGA reactors, none of the CHF correlations considered can be assumed to provide the definitive CHF power. It is recommended, however, to compare the power levels of the potential limiting rods with the power levels at which the Bernath and 2006 Groeneveld CHF correlations predict CHF to occur.
Date: March 30, 2008
Creator: Feldman, E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Maximum allowable heat flux for a submerged horizontal tube bundle

Description: For application to industrial heating of large pools by immersed heat exchangers, the so called maximum allowable (or critical) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. This is the condition at which vapor blanketing is expected to be initiated. Phenomenological considerations demonstrate why the maximum allowable heat flux would be expected to be less than for single tubes. Hydrodynamic theory is applied to extend the results of Lienhard and Dhir to large submerged bundles and the consequent correlation is compared to the correlation of Palen and Small and the limited data available for saturated conditions. To date the main conclusion is that estimates of q{double_prime}{sub chf} are highly uncertain for this configuration.
Date: December 31, 1996
Creator: McEligot, D. M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1994

Description: The physical phenomenon of forced convective boiling is probably one of the most interesting and complex transport phenomena. It has been under study for more than two centuries. Simply stated, forced convective subcooled boiling involves a locally boiling fluid: (1) whose mean temperature is below its saturation temperature, and (2) that flows over a surface exposed uniformly or non-uniformly to a high heat flux (HHF). The objective of this work is to assess and/or improve the present ability to predict local axial heat transfer distributions in the subcooled flow boiling regime for the case of uniformly heated coolant channels. This requires an accurate and complete representation of the boiling curve up to the CHF. The present. results will be useful for both heat transfer research and industrial design applications. Future refinements may result in the application of the results to non-uniformly heated channels or other geometries, and other fluids. Several existing heat transfer models for uniformly heated channels were examined for: (1) accurate representation of the boiling curve, and (2) characterizing the local heat transfer coefficient under high heat flux (HHF) conditions. Comparisons with HHF data showed that major correlation modifications were needed in the subcooled partial nucleate boiling (SPNB) region. Since the slope of boiling curve in this region is important to assure continuity of the HHF trends into the fully developed boiling region and up to the critical heat flux, accurate characterization in the SPNB region is essential. Approximations for the asymptotic limits for the SPNB region have been obtained and have been used to develop an improved composite correlation. The developed correlation has been compared with 363 water data points. For the local heat transfer coefficient and wall temperature, the over-all percent standard deviations with respect to the data were 19% and 3%, respectively, for the high ...
Date: July 1, 2000
Creator: Boyd, Dr. Ronald D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Critical heat flux and boiling heat transfer to water in a 3-mm-diameter horizontal tube.

Description: Boiling of the coolant in an engine, by design or by circumstance, is limited by the critical heat flux phenomenon. As a first step in providing relevant engine design information, this study experimentally addressed both rate of boiling heat transfer and conditions at the critical point of water in a horizontal tube of 2.98 mm inside diameter and 0.9144 m heated length. Experiments were performed at system pressure of 203 kPa, mass fluxes in range of 50 to 200 kg/m{sup z}s, and inlet temperatures in range of ambient to 80 C. Experimental results and comparisons with predictive correlations are presented.
Date: December 4, 2000
Creator: Yu, W.; Wambsganss, M. W.; Hull, J. R. & France, D. M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Boiling Visualization and Critical Heat Flux Phenomena In Narrow Rectangular Gap

Description: An experimental study was performed to investifate the pool boling critical hear flux (CHF) on one-dimensional inclined rectangular channels with narrow gaps by changing the orientation of a copper test heater assembly. In a pool of saturated water at atmospheric pressure, the test parameters include the gap sizes of 1,2,5, and 10 mm, andthe surface orientation angles from the downward facing position (180 degrees) to the vertical position (90 degress) respectively.
Date: December 1, 2004
Creator: Kim, J. J.; Kim, Y. H.; Kim, S. J.; Noh, S. W.; Suh, K. Y.; Rempe, J. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

8. Innovative Technologies: Two-Phase Heat Transfer in Water-Based Nanofluids for Nuclear Applications Final Report

Description: Abstract Nanofluids are colloidal dispersions of nanoparticles in water. Many studies have reported very significant enhancement (up to 200%) of the Critical Heat Flux (CHF) in pool boiling of nanofluids (You et al. 2003, Vassallo et al. 2004, Bang and Chang 2005, Kim et al. 2006, Kim et al. 2007). These observations have generated considerable interest in nanofluids as potential coolants for more compact and efficient thermal management systems. Potential Light Water Reactor applications include the primary coolant, safety systems and severe accident management strategies, as reported in other papers (Buongiorno et al. 2008 and 2009). However, the situation of interest in reactor applications is often flow boiling, for which no nanofluid data have been reported so far. In this project we investigated the potential of nanofluids to enhance CHF in flow boiling. Subcooled flow boiling heat transfer and CHF experiments were performed with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (≤ 0.1 % by volume) at atmospheric pressure. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient (HTC) are similar (within 20%). The HTC increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. The CHF tests were conducted at 0.1 MPa and at three different mass fluxes (1500, 2000, 2500 kg/m2s) under subcooled conditions. The maximum CHF enhancement was 53%, 53% and 38% for alumina, zinc oxide and diamond, respectively, always obtained at the highest mass flux. A post-mortem analysis of the boiling surface reveals that its morphology is altered by deposition of the particles during nanofluids boiling. A confocal-microscopy-based examination of the test section revealed that nanoparticles deposition not only changes the number of micro-cavities on the surface, but also the surface wettability. A simple model was ...
Date: July 31, 2009
Creator: Buongiorno, Jacopo & Hu, Lin-wen
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Two-Phase Spray Cooling with HFC-134a and HFO-1234yf for Thermal Management of Automotive Power Electronics using Practical Enhanced Surfaces

Description: The objective of this research was to investigate the performance of two-phase spray cooling with HFC-134a and HFO-1234yf refrigerants using practical enhanced heat transfer surfaces. Results of the study were expected to provide a quantitative spray cooling performance comparison with working fluids representing the current and next-generation mobile air conditioning refrigerants, and demonstrate the feasibility of this approach as an alternative active cooling technology for the thermal management of high heat flux power electronics (i.e., IGBTs) in electric-drive vehicles. Potential benefits of two-phase spray cooling include achieving more efficient and reliable operation, as well as compact and lightweight system design that would lead to cost reduction. The experimental work involved testing of four different enhanced boiling surfaces in comparison to a plain reference surface, using a commercial pressure-atomizing spray nozzle at a range of liquid flow rates for each refrigerant to determine the spray cooling performance with respect to heat transfer coefficient (HTC) and critical heat flux (CHF). The heater surfaces were prepared using dual-stage electroplating, brush coating, sanding, and particle blasting, all featuring "practical" room temperature processes that do not require specialized equipment. Based on the obtained results, HFC-134a provided a better heat transfer performance through higher HTC and CHF values compared to HFO-1234yf at all tested surfaces and flow rates. While majority of the tested surfaces provided comparable HTC and modestly higher CHF values compared to the reference surface, one of the enhanced surfaces offered significant heat transfer enhancement.
Date: August 2017
Creator: Altalidi, Sulaiman Saleh
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries

An evaluation of enhanced cooling techniques for high-heat load absorbers.

Description: Many components of the storage ring and front ends in the third generation of light sources are subjected to high heat loads from intense x-rays. Temperature rises and thermal stresses in these components must be kept within acceptable limits of critical heat flux and low-cycle fatigue failure. One of the design solutions is to improve heat transfer to the cooling water either by increasing water velocity in the cooling channels or by using inserts, such as porous media, twisted tapes and wire springs. In this paper we present experimental and analytical results to compare various enhanced cooling techniques for conditions specific to heating from an x-ray fan.
Date: October 28, 2002
Creator: Sharma, S.; Doose, C.; Rotela, E. & Barickowski, A.
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Partner: UNT Libraries Government Documents Department

Mechanistic modeling of CHF in forced-convection subcooled boiling

Description: Because of the complexity of phenomena governing boiling heat transfer, the approach to solve practical problems has traditionally been based on experimental correlations rather than mechanistic models. The recent progress in computational fluid dynamics (CFD), combined with improved experimental techniques in two-phase flow and heat transfer, makes the use of rigorous physically-based models a realistic alternative to the current simplistic phenomenological approach. The objective of this paper is to present a new CFD model for critical heat flux (CHF) in low quality (in particular, in subcooled boiling) forced-convection flows in heated channels.
Date: May 1, 1997
Creator: Podowski, M.Z.; Alajbegovic, A.; Kurul, N.; Drew, D.A. & Lahey, R.T. Jr.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

Description: Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.
Date: September 1, 1997
Creator: Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E. Kubik, D.L. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1996

Description: For the past decade, efforts have been growing in the development of high heat flux (HHF) components for many applications, including fusion and fission reactor components, advanced electronic components, synchrotrons and optical components, and other advanced HHF engineering applications. From a thermal prospective, work in the fusion reactor development arena has been underway in a number of areas including: (1) Plasma thermal, and electro-magnetics, and particle transport, (2) Fusion material, rheology, development, and expansion and selection; (3) High heat flux removal; and (4) Energy production and efficiency.
Date: July 1, 2000
Creator: Boyd, Dr. Ronald D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1997

Description: The Thermal Science Research Center (TSRC) at Prairie View A&M University is involved in an international fusion reactor technology development program aimed at demonstrating the technical feasibility of magnetic fusion energy. This report highlights: (1) Recent accomplishments and pinpoints thermal hydraulic problem areas of immediate concern to the development of plasma-facing components, and (2) Next generation thermal hydraulic problems which must be addressed to insure safety and reliability in component operation. More specifically, the near-term thermal hydraulic problem entails: (1) generating an appropriate data base to insure the development of single-side heat flux correlations, and (2) evaluating previously developed single-side/uniform heated transformations and correlations to determine which can be used to relate the vast two-phase heat transfer and critical heat flux (CHF) technical literature for uniformly heated flow channels to single-side heated channels.
Date: July 1, 2000
Creator: Boyd, Dr. Ronald D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Maximum allowable heat flux for a submerged horizontal tube bundle

Description: For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or {open_quotes}critical{close_quotes}) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration.
Date: August 14, 1995
Creator: McEligot, D.M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

RELAP5/MOD3.2 Assessment Using CHF Data from the KS-1 and V-200 Experiment Facilities

Description: The RELAP/MOD3.2 computer code has been assessed using rod bundle critical heat flux data from the KS-1 and V-200 facilities. This work was performed as part of the U.S. Department of Energy’s International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem 7, these tests addressed one of the important phenomena related to VVER behavior that the code needs to simulate well, core heat transfer. The code was judged to be in minimal agreement with the experiment data, consistently overpredicting the measured critical heat flux. It is recommended that a model development effort be undertaken to develop a critical heat flux model for RELAP5 that better represents the behavior in VVER rod bundles.
Date: July 1, 2001
Creator: Bayless, Paul David
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Pressure drop, heat transfer, critical heat flux, and flow stability of two-phase flow boiling of water and ethylene glycol/water mixtures - final report for project "Efficent cooling in engines with nucleate boiling."

Description: Because of its order-of-magnitude higher heat transfer rates, there is interest in using controllable two-phase nucleate boiling instead of conventional single-phase forced convection in vehicular cooling systems to remove ever increasing heat loads and to eliminate potential hot spots in engines. However, the fundamental understanding of flow boiling mechanisms of a 50/50 ethylene glycol/water mixture under engineering application conditions is still limited. In addition, it is impractical to precisely maintain the volume concentration ratio of the ethylene glycol/water mixture coolant at 50/50. Therefore, any investigation into engine coolant characteristics should include a range of volume concentration ratios around the nominal 50/50 mark. In this study, the forced convective boiling heat transfer of distilled water and ethylene glycol/water mixtures with volume concentration ratios of 40/60, 50/50, and 60/40 in a 2.98-mm-inner-diameter circular tube has been investigated in both the horizontal flow and the vertical flow. The two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux of the test fluids were determined experimentally over a range of the mass flux, the vapor mass quality, and the inlet subcooling through a new boiling data reduction procedure that allowed the analytical calculation of the fluid boiling temperatures along the experimental test section by applying the ideal mixture assumption and the equilibrium assumption along with Raoult's law. Based on the experimental data, predictive methods for the two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux under engine application conditions were developed. The results summarized in this final project report provide the necessary information for designing and implementing nucleate-boiling vehicular cooling systems.
Date: January 19, 2011
Creator: Yu, W.; France, D. M. & Routbort, J. L. (Energy Systems)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Critical heat flux tests with high pressure water in an internally heated annulus with alternating axial heat flux distribution

Description: Critical heat flux experiments were performed with an alternating heat flux profile in an internally heated annulus. The heated length was 84 inches with a square wave alternating heat flux profile over the last 12 inches having a maximum-to-average heat flux ratio of 1.76. Test data were obtained at pressures from 800 to 2000 psia, mass velocities from 0.25 x 10/sup 6/ to 2.8 x 10/sup 6/ lb/hr-ft/sup 2/ and inlet temperatures ranging from 400 to 600/sup 0/F. Two different electrically heated test sections were employed both with 72 inch uniform and 12 inch alternating heat flux sections. The second test section had a 0.44 inch hot patch with a peak-to-average heat flux ratio of 2.7 superimposed on the alternating flux profile at the exit end. Critical heat flux results with the alternating heat flux profile and with the superimposed hot patch were shown to be equivalent to those obtained in previous tests with a uniform heat flux profile except for several data points at low mass velocity and high enthalpy for which there is an apparent experimental bias in the uniform heat flux results.
Date: September 1, 1979
Creator: Beus, S.G. & Humphreys, D.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Static flow instability in subcooled flow boiling in parallel channels

Description: A series of tests for static flow instability or flow excursion (FE) at conditions applicable to the proposed Advanced Neutron Source reactor was completed in parallel rectangular channels configuration with light water flowing vertically upward at very high velocities. True critical heat flux experiments under similar conditions were also conducted. The FE data reported in this study considerably extend the velocity range of data presently available worldwide. Out of the three correlations compared, the Saha and Zuber correlation had the best fit with the data. However, a modification was necessary to take into account the demonstrated dependence of the Stanton (St) and Nusselt (Nu) numbers on subcooling levels, especially in the low subcooling regime.
Date: April 1, 1995
Creator: Siman-Tov, M.; Felde, D.K.; McDuffee, J.L. & Yoder, G.L. Jr.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department