Critical heat flux evaluation

International Nuclear Information System (INIS)

Banner, D.

1995-01-01

Critical heat flux (CHF) is of importance for nuclear safety and represents the major limiting factors for reactor cores. Critical heat flux is caused by a sharp reduction in the heat transfer coefficient located at the outer surface of fuel rods. Safety requires that this phenomenon also called the boiling crisis should be precluded under nominal or incidental conditions (Class I and II events). CHF evaluation in reactor cores is basically a two-step approach. Fuel assemblies are first tested in experimental loops in order to determine CHF limits under various flow conditions. Then, core thermal-hydraulic calculations are performed for safety evaluation. The paper will go into more details about the boiling crisis in order to pinpoint complexity and lack of fundamental understanding in many areas. Experimental test sections needed to collect data over wide thermal-hydraulic and geometric ranges are described CHF safety margin evaluation in reactors cores is discussed by presenting how uncertainties are mentioned. From basic considerations to current concerns, the following topics are discussed; knowledge of the boiling crisis, CHF predictors, and advances thermal-hydraulic codes. (authors). 15 refs., 4 figs

Prediction of critical heat flux in vertical pipe flow

International Nuclear Information System (INIS)

Levy, S.; Healzer, J.M.; Abdollahian, D.

1981-01-01

A previously developed semi-empirical model for adiabatic two-phase annular flow ix extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients. (orig.)

Measurements of Critical Heat Flux using Mass Transfer System

Energy Technology Data Exchange (ETDEWEB)

Hong, Seung Hyun; Chung Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

2016-05-15

In a severe accident, the reactor vessel is heated by the decay heat from core melts and the outer surface of reactor vessel is cooled by the natural convection of water pool. When the heat flux increases, boiling will start. Further increase of the heat flux may result in the CHF, which is generated by the bubble combinations. The CHF means that the reactor vessel was separated with coolant and wall temperature is raised rapidly. It may damage the reactor vessel. Also the CHF indicates the maximum cooling capability of the system. Therefore, the CHF has been used as a criterion for the regulatory and licensing. Mechanism of hydrogen vapor bubbles generated and combined can be simulated water bubbles mechanism. And also the both heat and mass transfer mechanism of CHF can be identified in the same methods. Therefore, the CHF phenomena can be simulated enough by mass transfer.

Research on cooling of ultra high critical heat flux with external flow boiling of water. Challenge to achieve ultra high critical heat flux and improvement in estimation of critical heat flux. JAERI's nuclear research promotion program, H11-004 (Contract research)

International Nuclear Information System (INIS)

Monde, Masanori; Mitsutake, Yuichi; Ishida, Kenji; Hino, Ryutaro

2003-03-01

An ultra high critical heat flux (CHF) has been challenged with a highly subcooled water jet impinging on a small rectangular heated surface. Major objective of the study is to achieve an ultra high heat flux cooling as large as 100 MW/m 2 and to establish an accurate estimation method of the CHF. The experiments were carried out over the experimental range; a fixed jet diameter of 2 mm, jet velocity of 5 - 35 m/s, degree of subcooling of 80 - 170 K and system pressure of 0.1 - 1.0 MPa. The rectangular heated surface with a thin nickel foil of 0.03 - 0.3 mm in thickness, 5 and 10 mm in length, and 4 mm in width and heated by a direct current. Effects of thickness of heater wall, jet velocity and subcooling on the CHF were experimentally elucidated. The experimental results show that the CHF decreases about 50% as the heater thickness, namely heat capacity of heater decreases. Characteristics of the CHF with heater length of 10 mm are correlated within ±20% by the generalized correlation of subcooled CHF proposed by the authors. However, the CHF with the shorter heater length of 5 mm shows large deviation of -40% especially at lower subcooling and higher velocity. The maximum CHF of 212 MW/m 2 was achieved at the subcooling of 151 K, the jet velocity of 35 m/s and system pressure of 0.5 MPa. The maximum CHF under atmospheric pressure approaches to 48% of the ultimate maximum heat flux given by the assumptions that vapor molecules leave a liquid-vapor interface at the average speed of a Boltzman-Maxwellian gas and any molecules returning to the interface are not permitted. The ratio of the CHF and ultimate maximum heat flux was considerably enhanced from the existing record of 30%. This study can give the feasibility of ultra high heat flux removal facing in a development of components such as a diverter of a fusion reactor. (author)

Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Wu, Qiao [Oregon State Univ., Corvallis, OR (United States)

2015-04-30

This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

A technical basis for the flux corrected local conditions critical heat flux correlation

International Nuclear Information System (INIS)

Luxat, J.C.

2008-01-01

The so-called 'flux-corrected' local conditions CHF correlation was developed at Ontario Hydro in the 1980's and was demonstrated to successfully correlate the Onset of Intermittent Dryout (OID) CHF data for 37-element fuel with a downstream-skewed axial heat flux distribution. However, because the heat flux correction factor appeared to be an ad-hoc, albeit a successful modifying factor in the correlation, there was reluctance to accept the correlation more generally. This paper presents a thermalhydraulic basis, derived from two-phase flow considerations, that supports the appropriateness of the heat flux correction as a local effects modifying factor. (author)

Predicting critical heat flux in slug flow regime of uniformly heated ...

African Journals Online (AJOL)

Numerical computation code (PWR-DNBP) has been developed to predict Critical Heat Flux (CHF) of forced convective flow of water in a vertical heated channel. The code was based on the liquid sub-layer model, with the assumption that CHF occurred when the liquid film thickness between the heated surface and vapour ...

Effect of heated length on the Critical Heat Flux of subcooled flow boiling. 2. Effective heated length under axially nonuniform heating condition

International Nuclear Information System (INIS)

Kinoshita, Hidetaka; Yoshida, Takuya; Nariai, Hideki; Inasaka, Fujio

1998-01-01

Effect of heated length on the Critical Heat Flux (CHF) of subcooled flow boiling with water was experimentally investigated by using direct current heated tube made of stainless steel a part of whose wall thickness was axially cut for realizing nonuniform heat flux condition. The higher enhancement of the CHF was derived for shorter tube length. The effective heated length was determined for the tube under axially nonuniform heat flux condition. When the lower heat flux part below the Net Vapor Generation (NVG) heat flux exists at the middle of tube length, then the effective heated length becomes the tube length downstream the lower heat flux parts. However, when the lower heat flux part is above the NVG, then the effective heated length is full tube length. (author)

Assessment of TRACE code against CHF experiments

International Nuclear Information System (INIS)

Audrius Jasiulevicius; Rafael Macian-Juan; Paul Coddington

2005-01-01

Full text of publication follows: This paper reports on the validation of the USNRC 'consolidate' code TRACE with data obtained during Critical Heat Flux (CHF) experiments in single channels and round and annular tubes. CHF is one of the key reactor safety parameters, because it determines the conditions for the onset of transition boiling in the core rod bundles, leading to the low heat transfer rates characteristics of the post-CHF heat transfer regime. In the context of the participation of PSI in the the International Programme for uncertainty analysis BEMUSE, we have carried out extensive work for the validation of some important TRACE models. The present work is aimed at assessing the range of validity for the CHF correlations and post-CHF heat transfer models currently included in TRACE. The heat transfer experiments selected for the assessment were performed at the Royal Institute of Technology (RIT) in Stockholm, Sweden and at the Atomic Energy Establishment in Winfrith, UK. The experimental investigations of the CHF and post-CHF heat transfer at RIT for flow of water in vertical tubes and annulus were performed at pressures ranging from 1 to 20 MPa and coolant mass fluxes from 500 to 3000 kg/m 2 s. The liquid was subcooled by 10 deg. C and 40 deg. C at the inlet of the test section. The experiments were performed on two different types of test sections. Experiments with uniformly heated single 7.0 m long tubes were carried out with three different inner tube diameters of 10, 14.9 and 24.7 mm. A series of experiments with non-uniform axial power distribution were also conducted in order to study the effect of the axial heat flux distribution on the CHF conditions in both 7.0 m long single tubes and 3.65 long annulus. Several different axial power profiles were employed with bottom, middle and top power peaks as well as the double-humped axial power profiles. In total more than 100 experiments with uniform axial heat flux distribution and several hundreds

Analysis of heat transfer under high heat flux nucleate boiling conditions

Energy Technology Data Exchange (ETDEWEB)

Liu, Y.; Dinh, N. [3145 Burlington Laboratories, Raleigh, NC (United States)

2016-07-15

Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

Analysis of heat transfer under high heat flux nucleate boiling conditions

International Nuclear Information System (INIS)

Liu, Y.; Dinh, N.

2016-01-01

Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

Simulation of CHF Condition using an Electroplating System

Energy Technology Data Exchange (ETDEWEB)

Ohk, Seung-Min; Park, Hae-Kyun; Chung, Bum-Jin [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

Heat transfer is enhanced when the bubbles are generated on the heated surface at the nucleate boiling regime since vigorous mixing of the liquid occurs near the heated surface due to the buoyancy force of the bubbles. As this phenomenon intensified, vapor film can be formed on the heated surface and it impairs heat transfer disturbing the heat exchange between the surface and the bulk liquid. And thus, the heat flux has the certain maximum value. This maximum value, Critical Heat Flux (CHF) is generally exhibits in the pool boiling condition in non-film boiling mode. Actually, the higher heat flux could be generated at the film boiling mode with extremely high surface temperature, which may unendurable for the system structure. CHF phenomena is simulated by hydrogen gas using electroplating system in mass transfer experiment. Vapor behavior on mass transfer experiment was visualized, and it was similar to that of on the heat transfer. CHF value was simulated by hydrogen gas with isovolumetric concept. Thus, virtual heat flux was estimated by mass flux, which is a non-heating process. Difference of gas density from heat transfer and mass transfer systems were considered and revised for the simulated heat flux. Despite of the simple parametric analysis, estimated CHF value of this study was 6.6 times smaller than Zuber's.

Critical heat flux correlation for thin rectangular channels

International Nuclear Information System (INIS)

Tanaka, Futoshi; Mishima, Kaichiro; Hibiki, Takashi

2007-01-01

The effect of heated length on Critical heat flux (CHF) in thin rectangular channels was studied based on CHF data obtained under atmospheric pressure. CHF in small channels has been widely studied in the past decades but most of the studies are related to CHF in round tubes. Although basic mechanisms of burnout in thin rectangular channels are similar to tubes, applicability of CHF correlations for tubes to rectangular channels are questionable since CHF in rectangular channels are affected by the existence of non-heated walls and the non-circular geometry of channel circumference. Several studies of CHF in thin rectangular channels have been reported in relation to thermal hydraulic design of research reactors and neutron source targets and CHF correlations have been proposed, but the studies mostly focus on CHFs under geometrical conditions of the application of interest. In his study, existing CHF data obtained in thin rectangular channels were collected and the effect of heated length on CHF was examined. Existing CHF correlations were verified with positive quality flow CHF data but none of the correlations successfully reproduced the CHF for a wide range of heated length. A new CHF correlation for qualify region applicable to a wide range of heated length was developed based on the collected data. (author)

Critical heat flux tests for a 12 finned-element assembly

Energy Technology Data Exchange (ETDEWEB)

Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

2017-03-15

Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

A theoretical critical heat flux model for low-pressure, low-mass-flux, and low-steam quality conditions

International Nuclear Information System (INIS)

Weihsiao Ho; Kuanchywan Tu; Baushei Pei; Chinjang Chang

1993-01-01

The critical heat flux (CHF) is the maximum heat flux just before a boiling crisis; its importance as a measurement of nuclear reactor power capability design as well as in the safety of reactors has been recognized. With emphasis on CHF behavior under subcooled and low-quality (i.e., 2 ·s), an improved model that uses the sublayer dry out theory has been developed. Based on experimental observations of CHF, the model assumes that CHF under such conditions is of the departure from nucleate boiling type. Based on the postulation that CHF is triggered by Helmholtz instability in the sublayer steam-liquid system, the model was developed by a simple energy balance of liquid sublayer evaporation as the vapor blanket tends to disturb the balance between the buoyancy force and the drag force exerted upon it. The model is compared with the well-known Biasi et al. correlation as well as the Atomic Energy of Canada Limited lookup table against 102 uniformly heated round tube CHF data and 34 nonuniformly heated round tube CHF data. The comparison shows that the model provides better accuracy and a reasonable agreement between the predicted values and experimental CHF data

An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

International Nuclear Information System (INIS)

Min Lee

1994-01-01

Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

Critical heat flux predictions for the Sandia Annular Core Research Reactor

International Nuclear Information System (INIS)

Rao, D.V.; El-Genk, M.S.

1994-08-01

This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

Inlet effect induced ''upstream'' critical heat flux in smooth tubes

International Nuclear Information System (INIS)

Kitto, J.B. Jr.

1986-01-01

An unusual form of ''upstream'' critical heat flux (CHF) has been observed and directly linked to the inlet flow pattern during an experimental study of high pressure (17 - 20 MPa) water flowing through a vertical 38.1 mm ID smooth bore tube with uniform axial and nonuniform circumferential heating. These upstream CHF data were characterized by temperature excursions which initially occurred at a relatively fixed axial location in the middle of the test section while the outlet and inlet heated lengths experienced no change. A rifled tube inlet flow conditioner could be substituted for a smooth tube section to generate the desired swirling inlet flow pattern. The upstream CHF data were found to match data from a uniformly heated smooth bore tube when the comparison was made using the peak local heat flux. The mechanism proposed to account for the upstream CHF observations involves the destructive interference between the decaying swirl flow and the secondary circumferential liquid flow field resulting from the one-sided heating

Critical heat flux concerns during the flow instability phase of a DEGB LOCA

International Nuclear Information System (INIS)

Shadday, M.A. Jr.

1990-08-01

Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding accident, a DEGB LOCA, the risk of CHF and attendant burnout is negligible. A review of RDAP data revealed that in the past reactor assemblies operated at flow and power conditions similar to those expected in a LOCA without burnout occurring. This is strong bounding empirical evidence, without the scaling concerns of laboratory experiments. A bounding analysis of the influences of assembly non-idealities on CHF, power tilts, and channel eccentricity, is included. The margin between operating heat fluxes, during the postulated LOCA, and CHF was quantified by scoping calculations. Based on measured azimuthal power variations, the local heat flux would have to be more than 20 standard deviations above the calculated mean heat flux for CHF to occur

Critical heat-flux experiments under low-flow conditions in a vertical annulus

International Nuclear Information System (INIS)

Mishima, K.; Ishii, M.

1982-03-01

An experimental study was performed on critical heat flux (CHF) at low flow conditions for low pressure steam-water upward flow in an annulus. The test section was transparent, therefore, visual observations of dryout as well as various instrumentations were made. The data indicated that a premature CHF occurred due to flow regime transition from churn-turbulent to annular flow. It is shown that the critical heat flux observed in the experiment is essentially similar to a flooding-limited burnout and the critical heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow regime transition. The observed CHF values are much smaller than the standard high quality CHF criteria at low flow, corresponding to the annular flow film dryout. This result is very significant, because the coolability of a heater surface at low flow rates can be drastically reduced by the occurrence of this mode of CHF

Anomalies and other concerns related to the critical heat flux

International Nuclear Information System (INIS)

Groeneveld, D.C.

2009-01-01

This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes. They include the following: Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated tubes. This phenomenon has been observed at high flows in both water and Freon. Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30 - 90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment-controlled and the deposition-controlled region and causes problems for prediction methods of the form CHF=f(X). Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, thus diminishing the effects of flow regime/heat transfer regime transitions responsible for the above unusual CHF trends. This will lead to a more gradually decreasing CHF vs. X curve. Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a dryout temperature excursion) to a positive slope. Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the shape of the axial temperature distribution of a heated tube experiencing CHF or heat transfer deterioration, when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. CHF fluid-to-fluid modeling: differences in CHF behavior at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for predicting CHF for bundle geometries

A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

International Nuclear Information System (INIS)

Yuan, Lan Qin; Yang, Jun; Harrison, Noel

2014-01-01

Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

Pool Boiling CHF in Inclined Narrow Annuli

International Nuclear Information System (INIS)

Kang, Myeong Gie

2010-01-01

Pool boiling heat transfer has been studied extensively since it is frequently encountered in various heat transfer equipment. Recently, it has been widely investigated in nuclear power plants for application to the advanced light water reactors designs. Through the review on the published results it can be concluded that knowledge on the combined effects of the surface orientation and a confined space on pool boiling heat transfer is of great practical importance and also of great academic interest. Fujita et al. investigated pool boiling heat transfer, from boiling inception to the critical heat flux (CHF, q' CHF ), in a confined narrow space between heated and unheated parallel rectangular plates. They identified that both the confined space and the surface orientation changed heat transfer much. Kim and Suh changed the surface orientation angles of a downward heating rectangular channel having a narrow gap from the downward-facing position (180 .deg.) to the vertical position (90 .deg.). They observed that the CHF generally decreased as the inclination angle (θ ) increased. Yao and Chang studied pool boiling heat transfer in a confined heat transfer for vertical narrow annuli with closed bottoms. They observed that when the gap size ( s ) of the annulus was decreased the effect of space confinement to boiling heat transfer increased. The CHF was occurred at much lower value for the confined space comparing to the unconfined pool boiling. Pool boiling heat transfer in narrow horizontal annular crevices was studied by Hung and Yao. They concluded that the CHF decreased with decreasing gap size of the annuli and described the importance of the thin film evaporation to explain the lower CHF of narrow crevices. The effect of the inclination angle on the CHF on countercurrent boiling in an inclined uniformly heated tube with closed bottoms was also studied by Liu et al. They concluded that the CHF reduced with the inclination angle decrease. A study was carried out

CHF correlations related to the core cooling of a research reactor

International Nuclear Information System (INIS)

Mishima, K.; Nishihara, H.; Shibata, T.

1984-01-01

Critical heat flux (CHF) at low flow condition can become important in a research reactor under a number of accident conditions. Regardless of the initial stages of these accidents, a similar condition, which is basically the decay heat removal by natural convection boiling, can develop. Under such conditions, burnout may occur even at a very low heat flux. In view of this, the low-flow CHF has been studied to provide a better understanding of the dryout behavior. The experimental results under atmospheric pressure indicate that a CHF can occur at much lower heat flux than pool-boiling CHF or than predicted by the conventional correlations. This fact indicates that a special care should be taken in analyzing the boiling phenomenon which occurs when the coolant flow is very low in a low pressure system. (author)

Anomalies and other concerns related to the critical heat flux

Energy Technology Data Exchange (ETDEWEB)

Groeneveld, D.C., E-mail: thermal@magma.ca [Researcher Emeritus, Chalk River Laboratories, Atomic Energy of Canada Ltd, Chalk River (Canada) and University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada)

2011-11-15

This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following: Bullet Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon. Bullet Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X). Bullet Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve. Bullet Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope. Bullet Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. Bullet CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for bundle geometries are

Ultrahigh Flux Thin Film Boiling Heat Transfer Through Nanoporous Membranes.

Science.gov (United States)

Wang, Qingyang; Chen, Renkun

2018-05-09

Phase change heat transfer is fundamentally important for thermal energy conversion and management, such as in electronics with power density over 1 kW/cm 2 . The critical heat flux (CHF) of phase change heat transfer, either evaporation or boiling, is limited by vapor flux from the liquid-vapor interface, known as the upper limit of heat flux. This limit could in theory be greater than 1 kW/cm 2 on a planar surface, but its experimental realization has remained elusive. Here, we utilized nanoporous membranes to realize a new "thin film boiling" regime that resulted in an unprecedentedly high CHF of over 1.2 kW/cm 2 on a planar surface, which is within a factor of 4 of the theoretical limit, and can be increased to a higher value if mechanical strength of the membranes can be improved (demonstrated with 1.85 kW/cm 2 CHF in this work). The liquid supply is achieved through a simple nanoporous membrane that supports the liquid film where its thickness automatically decreases as heat flux increases. The thin film configuration reduces the conductive thermal resistance, leads to high frequency bubble departure, and provides separate liquid-vapor pathways, therefore significantly enhances the heat transfer. Our work provides a new nanostructuring approach to achieve ultrahigh heat flux in phase change heat transfer and will benefit both theoretical understanding and application in thermal management of high power devices of boiling heat transfer.

Flow induced vibration characteristics in 2X3 bundle critical heat flux experiment

International Nuclear Information System (INIS)

Kim, Dae Hun; Chang, Soon Heung

2005-01-01

Above a certain heat flux, the liquid can no longer permanently wet the heater surface. This situation leads to an inordinate decrease in the surface heat transfer. This heat flux is commonly referred to as the critical heat flux (CHF). The CHF in nuclear reactors is one of the important thermal hydraulic parameters limiting the available power. Flow induced vibration (FIV) is the vibration caused by a fluid flowing around a body. In the fluid flowing system, FIV occurred by structures and flow condition. Many structures in nuclear power plant system are designed to prevent from structure failure due to FIV. Recently, Hibiki and Ishii (1998) carried out an experimental investigation on the effect of flow-induced vibration (FIV) on two-phase flow structure in vertical tube and reported that the FIV drastically changed the void fraction profiles. The void fraction profiles is one of the important parameter for determining CHF. Therefore, the investigation on the effect of FIV on CHF are needed. The research on FIV characteristics detection during CHF experiment in 2X3 bundle using R-134a has been carried out in KAIST. Using the results new FIV correlation in 2-pahse turbulent flow are suggested after finding out relation between CHF and dynamic pressure fluctuation value

Experimental and analytical studies of high heat flux components for fusion experimental reactor

International Nuclear Information System (INIS)

Araki, Masanori

1993-03-01

In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 ± 1 MW/m 2 was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate has been analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads. (J.P.N.) 62 refs

Critical heat flux for flow boiling of water in mini-channels

International Nuclear Information System (INIS)

Zhang, Weizhong; Mishima, Kaichiro; Hibiki, Takashi

2007-01-01

Critical heat flux (CHF) is a limiting factor when flow boiling is applied to dissipate high heat flux in mini-channels. In view of practical importance of critical heat flux correlations in engineering design and prediction, this study presents an evaluation of existing CHF correlations for flow boiling of water with available databases taken from small-diameter tubes, and then develops a new, simple CHF correlation for flow boiling in mini-channel. Three correlations by Bowring, Katto and Shah are evaluated with available CHF data in the literature for saturated flow boiling, and three correlations by Inasaka-Nariai, Celata et al. and Hall-Mudawar evaluated with the CHF data for subcooled flow boiling. The Hall-Mudawar correlation and the Shah correlation appear to be the most reliable tools for CHF prediction in the subcooled and saturated flow boiling regions, respectively. In order to avoid the defect of predictive discontinuities often encountered when applying previous correlations, a simple, nondimensional, inlet conditions dependent CHF correlation for saturated flow boiling has been formulated. Its functional form is determined by application of the artificial neural network and parametric trend analyses to the collected database. Superiority of this new correlation has been verified by the collected database. It has a mean deviation of 16.8% for this collected databank, smallest among all tested correlations. Compared to many inordinately complex correlations, this new correlation consists only of one single equation. (author)

Comparison of BIASI and Columbia CHF correlations using BODYFIT-2PE

International Nuclear Information System (INIS)

Chen, B.C.J.; Chien, T.H.; Sha, W.T.; Kim, J.H.

1984-01-01

This paper compares the BIASI critical heat flux (CHF) correlation with the Columbia CHF correlation by using both the homogeneous equilibrium two-phase model with algebraic slip and the drift flux model in BODYFIT-2PE. All calculations were compared with the GE 3 x 3 CHF experiment. This comparison serves as a qualification process for the CHF correlations in the framework of BODYFIT-2PE

Theoretical prediction method of subcooled flow boiling CHF

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min; Chang, Soon Heung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

Theoretical prediction method of subcooled flow boiling CHF

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min; Chang, Soon Heung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

International Nuclear Information System (INIS)

Williams, C.L.; Beus, S.G.

1980-05-01

Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values

Surface wettability effects on critical heat flux of boiling heat transfer using nanoparticle coatings

KAUST Repository

Hsu, Chin-Chi

2012-06-01

This study investigates the effects of surface wettability on pool boiling heat transfer. Nano-silica particle coatings were used to vary the wettability of the copper surface from superhydrophilic to superhydrophobic by modifying surface topography and chemistry. Experimental results show that critical heat flux (CHF) values are higher in the hydrophilic region. Conversely, CHF values are lower in the hydrophobic region. The experimental CHF data of the modified surface do not fit the classical models. Therefore, this study proposes a simple model to build the nexus between the surface wettability and the growth of bubbles on the heating surface. © 2012 Elsevier Ltd. All rights reserved.

Adaptation of a Freon-12 CHF correlation to apply for water in uniformly heated vertical tubes. Part 2: Based on CHF data for water at pressures in the range 6-20 MPa

International Nuclear Information System (INIS)

Green, W.J.

1982-03-01

An examination of more than 5000 sets of experimental data for critical heat flux (CHF) in uniformly heated vertical tubes internally cooled by high pressure water has shown that the CHF correlation proposed in Part 1 of this work is accurate for water at pressures up to approximately 17 MPa, provided that minor modifications are made to the Prandtl number index, and the saturation boiling length function. For pressures greater than 17 MPa, CHF values calculated from the correlation are increasingly lower than the experimental data, particularly at low saturation boiling length ratios ( -1 m -2 or thermal equilibrium exit qualities are less than 0.1

Post CHF heat transfer and quenching

International Nuclear Information System (INIS)

Nelson, R.A.; Condie, K.G.

1980-01-01

This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident (LOCA) or the FLECHT and Semiscale experiments. The mechanisms are also important to the blowdown phase of a LOCA or the recent Loss-of-Fluid Test (LOFT) experiments L2-2 and L2-3, which were 200% cold leg break transients. These LOFT experiments experienced total core quenching in the early part of the blowdown phase at high (1000 psia) pressures. The mechanisms are also important to certain pressurized water reactor (PWR) operational transients where the reactor may operate in the post-CHF regime for short periods of time. Accurate prediction of the post-CHF heat transfer including core quench during these transients is of prime importance to limit maximum cladding temperatures and prevent cladding deformation

Experimental Study on CHF in a Hemispherical Narrow Gap

International Nuclear Information System (INIS)

Jeong, J.H.; Park, R.J.; Kang, K.H.; Kim, S.B.; Kim, H.D.

1999-01-01

As a part of the SONATA-IV program, KAERI is conducting an experimental investigation of critical heat flux(CHF) in hemispherical narrow gaps. A visualization experiment, VISU-II, was done as the first step to get a visual observation of the flow behaviour inside a hemispherical gap and to understand the CHF-triggering mechanism. It was observed that the counter-current flow limitation (CCFL) phenomenon prevented water from wetting the heater surface and induced CHF. The CHFG (Critical Heat Flux in Gap) test is now being performed to measure the CHF and to investigate the inherent cooling mechanism in hemispherical narrow gaps. Temperature measurements over the heater surface show that the two-phase flow behaviour inside the gaps could be quite different from the other usual CHF experiments. The measured CHF points are lower than the predictions by existing empirical correlations based on the data measured with small-scale horizontal plates and vertical annulus. (authors)

Adaptation of a Freon-12 critical heat flux correlation to correlate water data from uniformly heated vertical tubes. Part I: Based on critical heat flux data for water at pressures of 3 to 14 MPa

International Nuclear Information System (INIS)

Green, W.J.

1981-12-01

Comparisons have been made between experimental critical heat flux (CHF) data for upflow of water in uniformly heated vertical tubes and values calculated from an empirical CHF correlation developed from Freon-12 data. When this correlation is re-evaluated to account for vapour Prandtl number effects, very good agreement is obtained between experimental data and calculated values over a wide range of coolant conditions. Comparison of values calculated from the revised correlation with 2063 sets of CHF data obtained from experiments with water in vertical, uniformly heated tubes shows a mean ratio of the calculated to experimental CHF of 0.82 and an r.m.s. error of 5.8 per cent for the following coolant conditions: (1) local pressure of 3.4 to 12 MPa; (2) mass flux greater than approx. 300 kg s -1 m -2 , and (3) thermal equilibrium value of exit quality greater than 0.1

An experimental study on the flow instabilities and critical heat flux under natural circulation

International Nuclear Information System (INIS)

Kim, Yun Il

1993-02-01

This study has been carried out to investigate the hydrodynamic stabilities of natural circulation and to analyze Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations and the average mass flux have been observed. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. The results show that the flow instability has strongly dependent on geometric conditions and operating parameters, the inlet temperature and the heat flux of test section. It was found that unstable region for the heat flux and the inlet temperature exists between the single-phase stable region of low heat and low inlet temperature and the two-phase stable region of very high heat flux and high inlet temperature. The CHF data from the natural and forced circulation experiments have been compared each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions were low flow less than 70 kg/m 2 s of water in vertical round tube with diameter of 0.008m at near atmospheric pressure. In this study, no difference in CHF values is observed between natural and fored circulation. Since low flow usually has the oscillation characteristic of relatively low amplitude and high frequency, the effect of the flow instabilities on the CHF seems to be negligible

Study on tube critical heat flux data treatment with artificial neural networks

International Nuclear Information System (INIS)

Han Lang; Shan Jianqiang

2005-01-01

Prediction of the Critical Heat Flux (CHF) are analyzed by Artificial Neural Networks (ANN) to a CHF database for upward flow of water in uniformly heated vertical round tubes. The analysis is performed with three viewpoints hypothesis, i.e. for fixed inlet condition, fixed exit condition and local condition. Half of 6941 from CHF database data is trained through ANN, the trained ANN predicts the total CHF data better than any other conventional correlations, showing RMS error of 6.6%, 10.39% and 21.39%, respectively. (author)

The 2005 CHF look-up table

International Nuclear Information System (INIS)

Groeneveld, D.C.; Vasic, A.Z.; Leung, L.K.H.; Durmayaz, A.; Shan, J.Q.; Yang, J.; Cheng, S.C.

2005-01-01

Full text of publication follows: CHF Look-up tables have been used widely for the prediction of the Critical Heat Flux (CHF) The CHF look-up table is basically a normalized data bank. The first CHF look-up table was constructed by Doroshchuk et al. (1975), using a limited database of 5 000 data points. This table, and all subsequent tables, contain normalized CHF values for a vertical 8 mm water-cooled tube for various pressures, mass fluxes and qualities. The CHF table development work has since been in progress at various institutions (e.g. CENG-Grenoble, University of Ottawa (UO), Ottawa, IPPE, Obninsk, and AECL, Chalk River) using an ever increasing data base. The 1995 CHF look-up table employs a data base containing about 30 000 CHF points and provides CHF values for an 8 mm ID, water-cooled tube, for 19 pressures, 20 mass fluxes, and 23 qualities. covering the full range of conditions of practical interest. The 2005 CHF LUT is an update to the 1995 LUT and addresses several concerns raised in the literature. The major improvements made are: - enhancement of the quality of the data base of the CHF look-up table (identify outliers, improve screening procedures); - increase in the data base by adding recently obtained data; - employment of greater subdivision of the look-up table by using smaller intervals in the independent parameters (pressure, mass flux and quality) at conditions where the variation in CHF is significant; - improvement of the smoothness of the CHF look-up table. This was done by the use of logarithmic functions for CHF, using optimum Spline functions etc. A discussion of the impact of these changes on the prediction accuracy and table smoothness is presented. It will be shown that the 2005 CHF look-up table is characterized by a significant improvement in accuracy and smoothness. [1] D. Groeneveld is the corresponding author. He is an Adjunct Professor at the University of Ottawa. (authors)

Assessment of CHF enhancement mechanisms in a curved, rectangular channel subjected to concave heating

International Nuclear Information System (INIS)

Sturgis, J.C.; Mudawar, I.

1999-01-01

An experimental study was undertaken to examine the enhancement in critical heat flux (CHF) provided by streamwise curvature. Curved and straight rectangular flow channels were fabricated with identical 5.0 x 2.5 mm cross sections and heated lengths of 101.6 mm in which the heat was applied to only one wall--the concave wall (32.3 mm radius) in the curved channel and a side wall in the straight. Tests were conducted using FC-72 liquid with mean inlet velocity and outlet subcooling of 0.25 to 10 m s -1 and 3 to 29 C, respectively. Centripetal acceleration for curved flow reached 315 times earth's gravitational acceleration. Critical heat flux was enhanced due to flow curvature at all conditions but the enhancement decreased with increasing subcooling. For near-saturated conditions, the enhancement was approximately 60% while for highly subcooled flow it was only 20%. The causes for the enhancement were identified as (1) increased pressure on the liquid-vapor interface at wetting fronts, (2) buoyancy forces and (3) increased subcooling at the concave wall. Flow visualization tests were conducted in transparent channels to explore the role of buoyancy forces in enhancing the critical heat flux. These forces were observed to remove vapor from the concave wall and distribute it throughout the cross section. Vapor removal was only effective at near-saturated conditions, yielding the observed substantial enhancement in CHF relative to the straight channel

Critical heat flux analysis and R and D for the design of the ITER divertor

International Nuclear Information System (INIS)

Raffray, A.R.; Chiocchio, S.; Merola, M.; Tivey, R.; Vieider, G.; Schlosser, J.; Driemeyer, D.; Escourbiac, F.; Grigoriev, S.; Youchison, D.

1999-01-01

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ∼10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R and D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R and D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ∼1.4 or more for all divertor components. (orig.)

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Energy Technology Data Exchange (ETDEWEB)

Wolf, S.; Holmes, D.H.

1977-04-01

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

Flow-Boiling Critical Heat Flux Experiments Performed in Reduced Gravity

Science.gov (United States)

Hasan, Mohammad M.; Mudawar, Issam

2005-01-01

Poor understanding of flow boiling in microgravity has recently emerged as a key obstacle to the development of many types of power generation and advanced life support systems intended for space exploration. The critical heat flux (CHF) is perhaps the most important thermal design parameter for boiling systems involving both heatflux-controlled devices and intense heat removal. Exceeding the CHF limit can lead to permanent damage, including physical burnout of the heat-dissipating device. The importance of the CHF limit creates an urgent need to develop predictive design tools to ensure both the safe and reliable operation of a two-phase thermal management system under the reduced-gravity (like that on the Moon and Mars) and microgravity environments of space. At present, very limited information is available on flow-boiling heat transfer and the CHF under these conditions.

Critical heat flux under zero flow conditions in a vertical 3 X 3 rod bundle with a non-uniform axial heat flux

Energy Technology Data Exchange (ETDEWEB)

Cho, Seok; Chun, Se Young; Moon, Sang Ki; Baek, Won Pil

2003-11-01

KAERI has performed an experimental study of water Critical Heat Flux (CHF) under zero flow conditions with a non-uniformly heated 3 by 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.5 to 15.0 MPa and inlet water subcooling enthalpies from 67.5 to 351.5 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 by 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a consistent parametric trend. The CHFs occur in the upper region of the heated section, but the vertical distances of the detected CHFs from the bottom of the heated section are reduced as the system pressure increases. Even though the effects of the inlet water subcooling enthalpies and system pressure in the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.

Development and Assessment of a Bundle Correction Method for CHF

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Chang, Soon Heung

1993-01-01

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first dericatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of GHE data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions. (Author)

Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-02-28

The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

Parametric trends analysis of the critical heat flux based on artificial neural networks

International Nuclear Information System (INIS)

Moon, S.K.; Baek, W.P.; Chang, S.H.

1996-01-01

Parametric trends of the critical heat flux (CHF) are analyzed by applying artificial neural networks (ANNs) to a CHF data base for upward flow of water in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. Katto's and Groeneveld et al. dimensionless parameters are used to train the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS errors of 8.9%, 13.1% and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local conditions hypothesis, respectively. The parametric trends of the CHF obtained from those trained ANNs show a general agreement with previous understanding. In addition, this study provides more comprehensive information and indicates interesting points for the effects of the tube diameter, the heated length, and the mass flux. It is expected that better understanding of the parametric trends is feasible with an extended data base. (orig.)

Critical heat flux in bottom heated two-phase thermosyphon. Improvement in critical heat flux due to concentric tube; Katan shuchu kanetsugata niso netsu syphon no genkai netsu ryusoku. Nijukan ni yoru genkai netsu ryusoku no kaizen

Energy Technology Data Exchange (ETDEWEB)

Monde, M.; Mitsutake, Y. [Saga University, Saga (Japan). Faculty of Science and Engineering

2000-02-25

An experiment has been carried out to elucidate the critical heat flux (CHF) of an open two-phase thermosyphon with a bottom heated chamber in which heat is absorbed by evaporation of liquid. Another objective is to enhance the CHF using a concentric-tube by which counter-current flow of vapor and liquid in the throat of the chamber can be controlled well. The CHF data are measured for the saturated liquid of R 113 at a different pressure and different configuration of concentric tubes. The CHF data without the inner tube are in good agreement with the existing correlation and analytical result. The CHF increases by as much as several times of the CHF without the inner tube with an increase in the inner tube diameter up to a certain diameter of the inner tube and then decreases continuously as the inner tube diameter approaches the outer tube diameter. The optimum diameter of inner tube exists at which the CHF is maximum. (author)

An experimental study of flow boiling chf with porous surface coatings and surfactant solutions

International Nuclear Information System (INIS)

Sarwar, Mohammad Sohail

2007-02-01

The boiling crisis or critical heat flux (CHF) phenomenon is an enormously studied topic of the boiling heat transfer. The great interest in the CHF is due to practical motives, since it is desirable to design an equipment (heat exchanger or boiler, etc) to operate at as high a heat flux as possible with optimum heat transfer rates but without the risk of physical burnout. This study consists of two parts of flow boiling CHF experiment: with porous surface coated tubes and by using surfactant solutions as working fluid. In first part, the effect of micro- and nano-porous inside surface coated vertical tubes on the CHF was determined for flow boiling of water in vertical round tubes at atmospheric pressure. CHF was measured for a smooth and three different coated tubes, at mass fluxes of 100∼300 kg/m 2 s and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Greater CHF enhancement was found with microporous coatings. Al 2 O 3 microporous coatings with particle size <10 μm and coating thickness of 50 μm showed the best CHF enhancement. The maximum increase in the CHF was about 25% for microporous Al 2 O 3 . A wettability test was performed to study the physical mechanism of increase of CHF with microporous coated surfaces and contact angle was measured for smooth and coated surfaces. Pressure drop measurements were also performed across the coated tubes using the DP-cell apparatus. In second part, surfactant effect on the CHF was determined for water flow boiling at atmospheric pressure in a closed loop filled with solution of tri-sodium phosphate (TSP, Na 3 PO 4 ·12H 2 O). The TSP is usually added to the containment sump water to adjust pH level during accident in nuclear power plants. The CHF was measured for four different surfactant solutions of water in vertical tubes, at different mass fluxes (100 ∼ 500 kg/m 2 s) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Surfactant solutions in the range of 0.05%∼0.2% at low mass

Experiments of Pool Boiling Performance (Boiling Heat Transfer and Critical Heat Flux) on Designed Micro-Structures

International Nuclear Information System (INIS)

Kim, Seol Ha; Kang, Jun Young; Lee, Gi Chol; Kiyofumia, Moriyama; Kim, Moo Hwan; Park, Hyun Sun

2015-01-01

In general, the evaluation of the boiling performance mainly focuses on two physical parameters: boiling heat transfer (BHT) and critical heat flux (CHF). In the nuclear power plants, both BHT and CHF contribute the nuclear system efficiency and safety, respectively. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on Pin-fin effect analysis. In terms of CHF, critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on the roughness ratio. The extended heat transfer area contributes the boiling heat transfer increase on the structured surface, and its quantitative analysis has been performed. In terms of CHF, the critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. We suggested a capillary limit to CHF delay for modeling capillary induced liquid inflow through microstructured surfaces. The critical size of the capillary limit on the prepared structured surface, determined by a model, could be reasonable explanation points for the experimental results (optimal size for CHF delay). The present experimental results also showed clearly the critical size (10 - 20 μm) for CHF delay, predicted by capillary limit analysis. This study provides fundamental insight into BHT and CHF enhancement of structured surfaces, and an optimal design guide for the required CHF and boiling heat-transfer performance. Finally, this study can contribute the basic understanding of the boiling on designed microstructure surface, and it also suggest the optimal micro scaled structured surface of boiling

CHF during flow rate, pressure and power transients in heated channels

International Nuclear Information System (INIS)

Celata, G.P.; Cumo, M.

1987-01-01

The behaviour of forced two-phase flows following inlet flow rate, pressure and power transients is presented here with reference to experiments performed with a R-12 loop. A circular duct, vertical test section (L = 2300 mm; D = 7.5 mm) instrumented with fluid (six) and wall (twelve) thermocouples has been employed. Transients have been carried out performing several values of flow decays (exponential decrease), depressurization rates (exponential decrease) and power inputs (step-wise increase). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast transients. Data analysis for a better theoretical prediction of CHF occurrence during transient conditions has been accomplished, and design correlations for critical heat flux and time-to-crisis predictions have been proposed for the different types of transients

RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions

International Nuclear Information System (INIS)

Ruggles, E.; Williams, P.T.

1990-01-01

Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons

Azimuthal critical heat flux in narrow rectangular channels

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Hoon; Noh, Sang Woo; Kim, Sung Joong; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

2003-07-01

Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10mm, and the surface orientation angles from the downward-facing position (180{sup o}) to the vertical position (90{sup o}), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at the atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle above which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis.

Development of low flow critical heat flux correlation for HANARO

International Nuclear Information System (INIS)

Park, Cheol; Chae, Hee Taek; Hang, Gee Yang.

1997-07-01

A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs

Development of low flow critical heat flux correlation for HANARO

Energy Technology Data Exchange (ETDEWEB)

Park, Cheol; Chae, Hee Taek; Hang, Gee Yang

1997-07-01

A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs.

Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

2017-06-15

Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

Prediction of critical heat flux in narrow rectangular channels using an artificial neural network

International Nuclear Information System (INIS)

Zhou Lei; Yan Xiao; Huang Yanping; Xiao Zejun; Yu Jiyang

2011-01-01

The concept of Critical heat flux (CHF) and its importance are introduced and the meaning to research CHF in narrow rectangular channels independently is emphasized. This paper is the first effort to predict CHF in NRCs using aritificial neural network. The mathematical structure of the artificial neural network and the error back-propagation algorithm are introduced. To predict CHF, the four dimensionless groups are inputted to the neural network and the output is the dimensionless CHF. As the hidden nodes increased, the training error decreases while the testing error decreases firstly and then transition occurs. Based on this, the hidden nodes are set as 5 and the trained network predicts all of the training and testing data points with RMS=0.0016 and μ=1.0003, which is better than several well-known existing correlations. Based on the trained network, the effect of several parameters on CHF are simulated and discussed. CHF increases almost linearly as the inlet subcooling increases. And larger mass flux enhances the effect of the inlet subcooling. CHF increases with the mass flux increasing. And the effect seems to be a little stronger for relatively low system pressure. CHF decreases almost linearly as the system pressure increases for the fixed inlet condition. The slope of the curve also increases with higher mass flux. This observation is limited to the ranges of the experimental database. CHF decreases as the heated length is increased and the gradients of the curves become very sharp for relatively short channel. CHF increases slightly with the diameter increasing with the variance of the gap limited within 1 to 3 mm. For relatively low mass flux, the effect of the equivalent diameter on CHF is insignificant. As the width of the channel is large enough, the effect of the gap is quite the same as that of the equivalent diameter. A BPNN is successfully trained based on near 500 CHF data points in NRCs, which has much better performances than the

Influence of stiffness on CHF for horizontal tubes under LPLF conditions

Energy Technology Data Exchange (ETDEWEB)

Baburajan, P.K. [Nuclear Safety Analysis Division, AERB, Niyamak Bhavan, 400094 (India); Bisht, Govind Singh [Department of Mechanical Engineering, IIT Bombay, 400076 (India); Gaikwad, Avinash J. [Nuclear Safety Analysis Division, AERB, Niyamak Bhavan, 400094 (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, IIT Bombay, 400076 (India)

2014-10-01

Highlights: • Effect of stiffness on the CHF in horizontal tube under LPLF conditions is studied. • CHF increases with the increase in stiffness. • Correlation for the prediction of CHF as a function of stiffness is developed. • Correlation for mass flux at CHF in terms of stiffness and initial mass flux is given. • RELAP5 is capable of predicting the effect of stiffness on CHF. - Abstract: Studies reported in the past on critical heat flux (CHF) are mostly limited to vertical flow, large channel diameter, high pressure and high mass flux. Since horizontal flow is commonly encountered in boiler tubes, refrigerating equipments and nuclear reactor fuel channels (PHWR), there is a need to understand horizontal flow CHF, generate sufficient experimental database and to develop reliable predictive method. Few studies are reported on the effect of upstream flow restrictions on flow instabilities and CHF. The present work investigates the effect of upstream flow restriction on CHF in horizontal flow at near atmospheric pressure conditions. In the present study, stiffness is defined as the ratio of upstream flow restriction pressure drop to the test section pressure drop. The classification of a flow boiling system as soft or stiff on the basis of quantification of the stiffness is attempted. Experimental data shows an increase in the CHF with the increase in the stiffness for a given initial mass flux. A correlation for the prediction of CHF under various stiffness conditions is developed. A correlation is suggested to predict the mass flux at CHF as a function of stiffness and initial mass flux. Modeling and transient analysis of the stiffness effect on CHF is carried out using the thermal hydraulic system code RELAP5. The predicted phenomena are in agreement with the experimental observations.

Critical heat flux data in a vertical tube at low and medium pressures

Energy Technology Data Exchange (ETDEWEB)

Teyssedou, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Olekhnowitch, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Tapucu, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Champagne, P [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Groeneveld, D [Chalk River Laboratories, AECL Research, Chalk River (Canada)

1994-09-01

AECL Research and Ecole Polytechnique have been cooperating on the validation of the critical heat flux (CHF) look-up table (D.C. Groeneveld et al., Heat Transfer Eng. 7(1-2) (1986) 46-62). For low and medium pressures the values in the table have been obtained by extrapolation and curve fitting; therefore, errors could be expected. To reduce these possible extrapolation errors, CHF experiments are being carried out in water cooled 8mm internal diameter (ID) tubes, at conditions where the data are scarce. This paper presents some of the experimental CHF data obtained for vertical up flow in an 8mm ID test section, for a wide range of exit qualities (5-70%) and the exit pressure ranging from 5 to 30bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8m. In general, the collected data show parametric trends similar to those described in the open literature. However, it was observed that for low pressure conditions CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30bar. The CHF data have also been compared with predictions of well-known correlations (L. Biasi et al., Energia Nucl. 14(9) (1967) 530-536; R. Bowring, Br. Report AEEW-R789, Winfrith, UK, 1972; Y. Khatto and H. Ohno, Int. J. Heat Mass Transfer 27 (1984) 1641-1648) and those of the look-up table given by Groeneveld et al. For low pressures and low mass fluxes the look-up table seems to yield better predictions of the CHF than the correlations. However, for medium pressures and mass fluxes the correlations perform better than the look-up table; among those tested, Katto and Ohno's correlation gives the best results. ((orig.))

Using a thermalhydraulics system code to estimate heat transfer coefficients for a critical heat flux experiment

International Nuclear Information System (INIS)

Statham, B.A.

2009-01-01

RELAP5/SCDAPSIM MOD 3.4 is used to predict wall temperature before and after critical heat flux (CHF) is reached in a vertical, uniformly heated tube using light water as the working fluid. The heated test section is modeled as a 1 m long Inconel 600 tube having an OD of 6.35 mm and ID of 4.57 mm with a 0.5 m long unheated development length at the inlet. Simulations are performed at pressures of 0.5 to 2.0 MPa with mass fluxes from 500 to 2000 kg m -2 s -1 and inlet qualities ranging from -0.2 to 0. Loss of flow simulations are performed with flow reduction rates of 10, 20, 50, and 100 kg m -2 s -2 . Inlet mass flux at CHF was nominally independent of rate in the model; this may or may not be realistic. (author)

Prediction of critical heat flux by a new local condition hypothesis

International Nuclear Information System (INIS)

Im, J. H.; Jun, K. D.; Sim, J. W.; Deng, Zhijian

1998-01-01

Critical Heat Flux(CHF) was predicted for uniformly heated vertical round tube by a new local condition hypothesis which incorporates a local true steam quality. This model successfully overcame the difficulties in predicted the subcooled and quality CHF by the thermodynamic equilibrium quality. The local true steam quality is a dependent variable of the thermodynamic equilibrium quality at the exit and the quality at the Onset of Significant Vaporization(OSV). The exit thermodynamic equilibrium quality was obtained from the heat balance, and the quality at OSV was obtained from the Saha-Zuber correlation. In the past CHF has been predicted by the experimental correlation based on local or non-local condition hypothesis. This preliminary study showed that all the available world data on uniform CHF could be predicted by the model based on the local condition hypothesis

In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

Energy Technology Data Exchange (ETDEWEB)

Groeneveld, D.C., E-mail: degroeneveld@gmail.com

2017-06-15

In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

The button effect of CANFLEX bundle on the critical heat flux and critical channel power

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E; Inch, W [Atomic Energy of Canada Limited, Ontario (Canada)

1998-12-31

A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

The button effect of CANFLEX bundle on the critical heat flux and critical channel power

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

1997-12-31

A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

Premature and stable critical heat flux for downward flow in a narrow rectangular channel

International Nuclear Information System (INIS)

Lee, Juhyung; Chang, Soon Heung; Jeong, Yong Hoon; Jo, Daeseong

2014-01-01

It has been recommended that RRs and MTRs be designed to have sufficient margins for CHF and the onset of FI as well, since unstable flow could leads to premature CHF under very low wall heat flux in comparison to stable CHF. Even the fact and previous studies, however, the understanding of relationship among FI, premature CHF and stable CHF is not sufficient to date. In this regards, subcooled flow boiling in a vertical rectangular channel was experimentally investigated to enhance the understanding of the CHF and the effect of the two-phase flow instability on it under low pressure conditions, especially for downward flow which was adopted for Jordan Research and Training Reactor (JRTR) and Kijang research reactor (KJRR) to achieve easier fuel and irradiation rig loading. In this study, CHF for downward flow of water under low pressure in narrow rectangular channel was experimentally investigated. For conditions such as downward flow, narrow rectangular channel and low pressure, it has been deduced from literature that flow instability could largely influence on triggering CHF at lower heat flux, i. e. premature CHF. Total 54 CHF data, which includes premature and stable data was obtained for various fluid conditions and system configurations including inlet stiffness. The upper and lower boundaries of CHF were newly proposed based on the experiment

Critical heat flux in vertical flows at low pressures; Flux de chaleur critique en ecoulements verticaux aux pressions faibles

Energy Technology Data Exchange (ETDEWEB)

Olekhnowitch, A [Ecole Polytechnique, Montreal, PQ (Canada)

1994-12-31

This paper presents some critical heat flux (CHF) data obtained for vertical upflow of water in an 8 mm test section, for exit pressures ranging from 5 to 30 bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8 m. In general, the collected data show trends similar to those described in the open literature. However, it was observed that for low pressures CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30 bar. The data have been compared with a look-up table and predictions of well known correlations. For low pressures and low mass fluxes, the look-up table seems to give better predictions, but for medium pressures and mass fluxes, the correlations perform better. 19 refs., 5 figs.

Assessment of fluid-to-fluid modelling of critical heat flux in horizontal 37-element bundle flows

International Nuclear Information System (INIS)

Yang, S.K.

2006-01-01

Fluid-to-fluid modelling laws of critical heat flux (CHF) available in the literature were reviewed. The applicability of the fluid-to-fluid modelling laws was assessed using available data ranging from low to high mass fluxes in horizontal 37-element bundles simulating a CANDU fuel string. Correlations consisting of dimensionless similarity groups were derived using modelling fluid data (Freon-12) to predict water CHF data in horizontal 37-element bundles with uniform and non-uniform axial-heat flux distribution (AFD). The results showed that at mass fluxes higher than ∼4,000 kg/m 2 s (water equivalent value), the vertical fluid-to-fluid modelling laws of Ahmad (1973) and Katto (1979) predict water CHF in horizontal 37-element bundles with non-uniform AFD with average errors of 1.4% and 3.0% and RMS errors of 5.9% and 6.1%, respectively. The Francois and Berthoud (2003) fluid-to-fluid modelling law predicts CHF in non-uniformly heated 37-element bundles in the horizontal orientation with an average error of 0.6% and an RMS error of 10.4% over the available range of 2,000 to 6,200 kg/m 2 s. (author)

Analytical prediction of CHF by FIDAS code based on three-fluid and film-dryout model

International Nuclear Information System (INIS)

Sugawara, Satoru

1990-01-01

Analytical prediction model of critical heat flux (CHF) has been developed on the basis of film dryout criterion due to droplets deposition and entrainment in annular mist flow. Critical heat flux in round tubes were analyzed by the Film Dryout Analysis Code in Subchannels (FIDAS) which is based on the three-fluid, three-field and newly developed film dryout model. Predictions by FIDAS were compared with the world-wide experimental data on CHF obtained in water and Freon for uniformly and non-uniformly heated tubes under vertical upward flow condition. Furthermore, CHF prediction capability of FIDAS was compared with those of other film dryout models for annular flow and Katto's CHF correlation. The predictions of FIDAS are in sufficient agreement with the experimental CHF data, and indicate better agreement than the other film dryout models and empirical correlation of Katto. (author)

Critical heat flux for APR1400 lower head vessel during a severe accident

International Nuclear Information System (INIS)

Noh, Sang W.; Suh, Kune Y.

2013-01-01

Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

The Heat Flux Analysis in an Annulus Narrows Gap With Initial Temperature Variations Using HeaTiNG-01 Test Section

International Nuclear Information System (INIS)

Mulya Juarsa; Efrizon Umar; Andhang Widi Harto

2009-01-01

An experiment to understand the complexity of boiling phenomena on a narrow gap, which has occurs in severe accident at TMI-2 NPP is necessary to be done in aimed to increase the understanding of accident management. The goal of research is to obtain a heat flux and critical heat flux (CHF) value during boiling heat transfer process in a narrow gap annulus. The method of research is experimental using HeaTiNG-01 test section. The experiment has been done with heating-up heated rod until a certain initial temperature, for this experiment, three initial temperature variations was decided at 650°C, 750°C dan 850°C. Then, a cooling process in heated rod by saturated water was recorded based on temperature data changes. Temperature data was used to calculate a value of heat flux and wall superheat temperature, until the results could be defined in boiling curve. The result of this research shows that, although the initial temperature of heated rod was different, the value of CHF is almost similar with CHF average 253.7 kW/m 2 with the changes of only 4.7%. The event of boiling in a narrow gap is not included pool boiling category based on the comparison of film boiling area of the experiment to Bromley correlations. (author)

Prediction of sodium critical heat flux (CHF) in annular channel using grey systems theory

International Nuclear Information System (INIS)

Zhou Tao; Su Guanghui; Zhang Weizhong; Qiu Suizheng; Jia Dounan

2001-01-01

Using grey systems theory and experimental data obtained from sodium boiling test loop in China, the grey mutual analysis of some parameters influencing sodium CHF is carried out, and the CHF values are predicted by GM(1, 1) model. The GM(1, h) model is established for CHF prediction, and the predicted CHF values are good agreement with the experimental data

Doubling of critical heat flux using a grapheme oxide nanofluid and its repeatabiltiy

International Nuclear Information System (INIS)

Moon, Sung Bo; Bang, In Cheol

2013-01-01

CHF(Critical Heat Flux : heat flux which makes dramatic increase of temperature on heater surface) is one of the most important phenomena in the thermal hydraulic system. High CHF makes more thermal margin of heat transfer. This makes high efficiency and safety of power plant especially in nuclear power plant. Much smaller danger can be concerned to public society like radioactive material leakage in the accidents. Graphene Oxide which can be deposited on the heater surface makes nano-scale structures with enhancing thermal limit of heater. Three major models of enhancing limit of heater have been concerned in many heat transfer studies. In this study, wettability that is about ability to wet on surface and thermal activity which is about thermal property of coated layer are concerned to analyze the mechanism of CHF enhancing. Also, chemical reduction of Graphene Oxide(GO) to Reduced Graphene Oxide(RGO) on the surface will be concerned with one reason of changing wettability of nano-scale structure on the heater surface. We used GO nanofluid 0.001 volume percent. Two models are compared to explain how CHF is enhanced. Results show wettability increased with slightly reduced GO and structure. And in thermal activity model, the most powerful term, thickness of layer, is too small to affect thermal activity. It has low ability to explain how GO nanofluid can enhance CHF

New Westinghouse correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids

International Nuclear Information System (INIS)

Motley, F.E.; Hill, K.W.; Cadek, F.F.; Shefcheck, J.

1976-07-01

A new critical heat flux (CHF) correlation, based on local fluid conditions, has been developed from Westinghouse rod bundle data. This correlation applies to both 0.422 inch and 0.374 inch rod O.D. geometries. It accounts for typical cell and thimble cell effects, uniform and non-uniform heat flux profiles, variations in rod heated length and in grid spacing. The correlation predicts CHF for 1147 data points with a sample mean and standard deviation of measured-to-predicted heat flux ratio of 1.0043 and 0.0873, respectively. It was concluded that to meet the reactor design criterion the minimum DNBR should be 1.17

An experimental study on the flow instabilities and critical heat flux under natural circulation

International Nuclear Information System (INIS)

Kim, Yun II; Chang, Soon Heung

2004-01-01

This study has been carried out to investigate the hydrodynamic stabilities and Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations has been observed and the average mass flux under flow oscillation have been measured. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. And the CHF data from low flow experiments, namely the natural and forced circulation, have been compared with each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions for the CHF experiments were a low flow of less than 70 kg/m 2 s of water in a vertical round tube with diameter of 0.008 m at near atmospheric pressure. (author)

A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

International Nuclear Information System (INIS)

Ha, Sang Jun; No, Hee Cheon

1997-01-01

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variation in pressure, tube diameter and length, mass flux and inlet subcooling

A new mechanistic model of critical heat flux in forced-convection subcooled boiling

International Nuclear Information System (INIS)

Alajbegovic, A.; Kurul, N.; Podowski, M.Z.; Drew, D.A.; Lahey, R.T. Jr.

1997-10-01

Because of its practical importance and various industrial applications, the process of subcooled flow boiling has attracted a lot of attention in the research community in the past. However, the existing models are primarily phenomenological and are based on correlating experimental data rather than on a first-principle analysis of the governing physical phenomena. Even though the mechanisms leading to critical heat flux (CHF) are very complex, the recent progress in the understanding of local phenomena of multiphase flow and heat transfer, combined with the development of mathematical models and advanced Computational Fluid Dynamics (CFD) methods, makes analytical predictions of CHF quite feasible. Various mechanisms leading to CHF in subcooled boiling have been investigated. A new model for the predictions of the onset of CHF has been developed. This new model has been coupled with the overall boiling channel model, numerically implemented in the CFX 4 computer code, tested and validated against the experimental data of Hino and Ueda. The predicted critical heat flux for various channel operating conditions shows good agreement with the measurements using the aforementioned closure laws for the various local phenomena governing nucleation and bubble departure from the wall. The observed differences are consistent with typical uncertainties associated with CHF data

Development of the heated length to diameter correction factor on critical heat flux using the artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Ho; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

With using artificial neural networks (ANNs), an analytical study related to the heated length effect on critical heat flux (CHF) has been carried out to make an improvement of the CHF prediction accuracy based on local condition correlations or table. It has been carried out to suggest a feasible criterion of the threshold length-to-diameter (L/D) value in which heated length could affect CHF. And within the criterion, a L/D correction factor has been developed through conventional regression. In order to validate the developed L/D correction factor, CHF experiments for various heated lengths have been carried out under low and intermediate pressure conditions. The developed threshold L/D correlation provides a new feasible criterion of L/D threshold value. The developed correction factor gives a reasonable accuracy for the original database, showing the error of -2.18% for average and 27.75% for RMS, and promising results for new experimental data. 7 refs., 12 figs., 1 tab. (Author)

Development of the heated length to diameter correction factor on critical heat flux using the artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Ho; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

With using artificial neural networks (ANNs), an analytical study related to the heated length effect on critical heat flux (CHF) has been carried out to make an improvement of the CHF prediction accuracy based on local condition correlations or table. It has been carried out to suggest a feasible criterion of the threshold length-to-diameter (L/D) value in which heated length could affect CHF. And within the criterion, a L/D correction factor has been developed through conventional regression. In order to validate the developed L/D correction factor, CHF experiments for various heated lengths have been carried out under low and intermediate pressure conditions. The developed threshold L/D correlation provides a new feasible criterion of L/D threshold value. The developed correction factor gives a reasonable accuracy for the original database, showing the error of -2.18% for average and 27.75% for RMS, and promising results for new experimental data. 7 refs., 12 figs., 1 tab. (Author)

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

International Nuclear Information System (INIS)

Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E.; Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

1997-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

1997-09-01

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.

Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1993

International Nuclear Information System (INIS)

Boyd, Ronald D.

2000-01-01

Subcooled flow boiling in heated coolant channels is an important heat transfer enhancement technique in the development of fusion reactor components, where high heat fluxes must be accommodated. As energy fluxes increase in magnitude, additional emphasis must be devoted to enhancing techniques such as sub cooling and enhanced surfaces. In addition to subcooling, other high heat flux alternatives such as high velocity helium and liquid metal cooling have been considered as serious contenders. Each technique has its advantages and disadvantages [1], which must be weighed as to reliability and reduced cost of fusion reactor components. Previous studies [2] have set the stage for the present work, which will concentrate on fundamental thermal hydraulic issues associated with the h-international Thermonuclear Experimental Reactor (ITER) and the Engineering Design Activity (EDA). This proposed work is intended to increase our understanding of high heat flux removal alternatives as well as our present capabilities by: (1) including single-side heating effects in models for local predictions of heat transfer and critical heat flux; (2) inspection of the US, Japanese, and other possible data sources for single-side heating, with the aim of exploring possible correlations for both CHF and local heat transfer; and (3) assessing the viability of various high heat flux removal techniques. The latter task includes: (a) sub-cooled water flow boiling with enhancements such as twisted tapes, and hypervapotrons, (b) high velocity helium cooling, and (c) other potential techniques such as liquid metal cooling. This assessment will increase our understanding of: (1) hypervapotron heat transfer via fins, flow recirculation, and flow oscillation, and (2) swirl flow. This progress report contains selective examples of ongoing work. Section II contains an extended abstract, which is part of and evolving technical paper on single-side f heating. Section III describes additional details

CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)

2016-08-01

Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

Occurrence of critical heat flux during blowdown with flow reversal

International Nuclear Information System (INIS)

Leung, J.C.M.

1976-04-01

A small-scale experiment using Freon-11 at 130 0 F and 65 psia in a well-instrumented transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 2 ft length. Inlet and exit void fractions were measured by a capacitance technique. Flow regime transition was observed with high speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F and 60 psig was found to greatly affect the steady-state subcooled boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the thermodynamic nonequilibrium conditions required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown. Critical heat flux did not occur immediately during the flow decay in an approximately 60 msec reversal period. The first or early CHF which occurred at about 400 msec was independent of the blowdown volume and did not propagate upward. An annular flow pattern appeared at the onset of this CHF which occurred only at the lower 8 in. of the heated zone

Critical heat flux of subcooled flow boiling in a narrow tube

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki; Shimura, Toshiya.

1986-01-01

The critical heat flux (CHF) of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the following conditions: tube inside diameter: 1 ∼ 3 mm, tube length: 10 ∼ 100 mm, and water mass velocity: 7000 - 20000 kg/(m 2 · s). The critical heat flux increases the shorter the tube length and the smaller the tube inside diameter, at the same water mass velocity and exit quality. Experimental data were compared with empirical correlations, such as the Griffel and Knoebel correlations for subcooled boiling at low pressure, the Tong correlation for subcooled boiling at high pressure, and the Katto correlation. The existence of two parameter regions was confirmed. The first is the low CHF region in which experimental data can be predicted well by Griffel and Knoebel correlations, and the second is the high CHF region in which experimental data are higher than the predictions by the above two correlations. (author)

Critical heat flux in tubes and tight hexagonal rod lattices

International Nuclear Information System (INIS)

Erbacher, F.J.; Cheng Xu; Zeggel, W.

1994-01-01

The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

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

International Nuclear Information System (INIS)

Park, Jae-Wook; Kim, Hong-Chae; Beak, Won-Pil; Chang, Soon Heung

1997-01-01

A series of critical heat flux (CHF) tests have been performed to provide a reliable set of CHF data for water flow in vertical round tubes at low pressure and low flow (LPLF) conditions. The range of experimental conditions is as follows: diameter 8, 10 mm; heated length 0.5, 1 m; pressure 2-9 bar, mass flux 50-200 kg/m 2 s; inlet subcooling 350, 450 kJ/kg. The observed parametric trends are generally consistent with the previous understanding except for the effects of system pressure and tube diameter. The pressure effect is small but very complicated; existing CHF correlations do not represent this parametric trend properly. CHF increases with the increase in diameter at fixed exit conditions, contrary to the general understanding. The artificial neural networks are applied to the round tube CHF data base at LPLF (P = 110-1100 kPa, G = 0-500 kg/m 2 s) conditions. The trained backpropagation networks (BPNs) predict CHF better than any other CHF correlations. Parametric trends of CHF based on the BPN for fixed inlet conditions generally agree well with our experimental results. (author)

Counter-current flow limited CHF in thin rectangular channels

International Nuclear Information System (INIS)

Cheng, L.Y.

1990-01-01

An analytical expression for counter-current-flow-limitation (CCFL) was used to predict critical heat flux (CHF) for downward flow in thin vertical rectangular channels which are prototypes of coolant channels in test and research nuclear reactors. Top flooding is the mechanism for counter-current flow limited CHF. The CCFL correlation also was used to determine the circulation and flooding-limited CHF. Good agreements were observed between the period the model predictions and data on the CHF for downflow. The minimum CHF for downflow is lower than the flooding-limited CHF and it is predicted to occur at a liquid flow rate higher than that at the flooding limit. 17 refs., 7 figs

Improvement of critical heat flux correlation for research reactors using plate-type fuel

International Nuclear Information System (INIS)

Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

1998-01-01

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as Loss of the primary coolant flow'. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study. (author)

Prediction of critical heat flux using ANFIS

Energy Technology Data Exchange (ETDEWEB)

Zaferanlouei, Salman, E-mail: zaferanlouei@gmail.co [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of); Rostamifard, Dariush; Setayeshi, Saeed [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of)

2010-06-15

The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

Prediction of critical heat flux using ANFIS

International Nuclear Information System (INIS)

Zaferanlouei, Salman; Rostamifard, Dariush; Setayeshi, Saeed

2010-01-01

The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

Energy Technology Data Exchange (ETDEWEB)

Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

1995-09-01

In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

Experimental study of CHF enhancement using Fe{sub 3}O{sub 4} nanofluids in the subcooled boiling region

Energy Technology Data Exchange (ETDEWEB)

Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-10-15

This study may give overall trends of CHF enhancement in the subcooled boiling region. In our experiment, subcooled flow boiling CHF enhancement phenomena in water and nano-coated surface was investigated in mass flux from 1000 to 5000 kg/m{sup 2}s. CHF enhancement of nanoparticles coated tube in DI water increased as exit quality get bigger at same massflux. Various methods to improve CHF characteristics are introduced, especially nanofluids are used for enhancing the CHF. Nanofluids is a colloidal suspension that nanoparticles are mixed with basic fluid. Normally the use of nanofluids as working fluid improves the flow boiling CHF characteristics. Lee et al. already researched the CHF characteristics using nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. CHF enhancement was observed when exit quality was low and a DNB-like thermal crisis occurred. But CHF enhancement didn't occur for high exit quality, but LFD-type thermal crisis occurred. Because LFD phenomena are nearly unaffected by the surface conditions, CHF enhancement is not expected for annular flow with high exit quality. Kim et al. performed flow boiling CHF enhancement at subcooled region using alumina-water, zinc-oxide-water and diamond-water nanofluids. The CHF was enhanced by increasing wettability from nanoparticle deposition. CHF enhancement occurred in high mass flux (2000-2500 kg/m{sup 2}s), but CHF enhancement didn't occur in low mass flux (1500 kg/m{sup 2}s). The amount of nanoparticle deposition on each tube can be different during experiments by the several conditions such as deposition time, mass flux and heat flux. So, before the nanofluid experiment conducted, all tube are deposited in same condition of heat flux, concentration and time.

A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

Energy Technology Data Exchange (ETDEWEB)

Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

Energy Technology Data Exchange (ETDEWEB)

Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

Experimental study of critical heat flux in inclined rectangular gap

International Nuclear Information System (INIS)

Kim, S.J.; Kim, Y.H.; Noh, S.W.; Suh, K.Y.; Rempe, J.L.; Cheung, F.B.; Kim, S.B.

2003-01-01

In the TMI-2 accident, the lower part of the reactor pressure vessel was overheated and then rather rapidly cooled down, as was later found out in a vessel investigation project. This accounted for the possibility of gap cooling feasibility. For this reason, a great deal of investigations was performed to determine the critical heat flux (CHF) from the standpoint of in-vessel retention (IVR). As part of a joint Korean-U.S. International Nuclear Energy Research Initiative (INERI) project, Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180deg) to the vertical position (90deg), respectively. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. However, in downward-facing position (180deg), somewhat differing results were detected relative to previous reports. For a certain gap size having a similar dimension with vapor layer thickness, more efficient heat transfer was detected and this may be interpreted by characteristic property such as the vapor layer thickness of water. In consistency with several studies reported in the literature, it was found that there exists a transition angle above that the CHF changes with a rapid slope. (author)

An investigation of critical heat fluxes in vertical tubes internally cooled by Freon-12. Part I - Critical heat flux experiments with axially uniform and non-uniform heating and comparisons of data with selected correlations

International Nuclear Information System (INIS)

Green, W.J.; Stevens, J.R.

1981-08-01

Experiments have been performed using vertical heated tubes, cooled internally by Freon-12, to determine critical heat fluxes (CHFs) for both a uniformly heated section and an exit region with a separately controlled power supply. Heated lengths of the main separately were 2870 mm (8.48 and 16.76 mm tube bores) and 3700 mm (for 21.34 mm tube bore); heated length of the exit section was 230 mm. Coolant pressures, exit qualities and mass fluxes were in the range 0.9 to 1.3 MPa, 0.19 to 0.86, and 380 to 2800 kg m -2 s -1 , respectively. The data have been compared with published empirical correlations specifically formulated to predict CHFs in Freon-cooled, vertical tubes; relevant published CHF data have also been compared with these correlations. These comparisons show that, even over the ranges of conditions for which the correlations were developed, predicted values are only accurate to within +-20 per cent. Moreover, as mass fluxes increase above 3500 kg m -2 s -1 , the modified Groeneveld correlation becomes increasingly inadequate, and the Bertoletti and modified Bertoletti correlations under-predict CHF values by increasing amounts. At mass fluxes below 750 kg m -2 s -1 the Bertoletti correlations exhibit increasing inaccuracy with a decrease in mass flux. For non-uniform heating, the correlations are at variance with the experimental data

Phenomenological modelling of CHF in annular flow in annuli using new models of droplet deposition and entrainment

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F., E-mail: g.hewitt@imperial.ac.uk [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

2016-08-15

Highlights: • A phenomenological model to predict the CHF for flows in annuli is described. • New correlations of droplet entrainment and deposition are used. • The present model has good predictive capability in predicting CHF in annuli. - Abstract: In this paper, we present a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in heated vertical annuli. In present model, a new set of correlations of droplet deposition and entrainment in annuli was used which were verified by comparison with the data of Moeck (1970) for developing liquid films in adiabatic annuli. In the results presented here, these new correlations have been used to predict 2249 independent data on critical heat flux (CHF) obtained both regarding internal heating of the rod as well as simultaneous heating of the rod and the outer tube in six heated vertical annuli under various mass flow rate, pressure and inlet quality and where the conditions were such that (as is most common) the CHF condition occurred in the annular flow regime. The comparisons between the calculated and measured CHFs showed that the present model has good predictive capability in predicting CHF.

Experimental investigation of pool boiling heat transfer and critical heat flux on a downward facing surface

International Nuclear Information System (INIS)

Gocmanac, M.; Luxat, J.C.

2012-01-01

A separate effects experimental study of heat transfer and Critical Heat Flux (CHF) on a downward facing plate in subcooled water pool boiling is described. Two geometries of downwards facing surfaces are studied. The first is termed the 'confined' study in which bubble motion is restricted to the heated surface. The second is termed the 'unconfined' study where individual bubbles are free to move along the heated surface and vent in any direction. The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The CHF as a function of angle of inclination of the surface is presented and is in good agreement with other experimental data from somewhat different test geometries. (author)

Correlation of critical heat flux data for uniform tubes

Energy Technology Data Exchange (ETDEWEB)

Jafri, T.; Dougherty, T.J.; Yang, B.W. [Columbia Univ., New York, NY (United States)

1995-09-01

A data base of more than 10,000 critical heat flux (CHF) data points has been compiled and analyzed. Two regimes of CHF are observed which will be referred to as the high CHF regime and the low CHF regime. In the high CHF regime, for pressures less than 110 bar, CHF (q{sub c}) is a determined by local conditions and is adequately represented by q{sub c} = (1.2/D{sup 1/2}) exp[-{gamma}(GX{sub t}){sup 1/2}] where the parameter {gamma} is an increasing function of pressure only, X{sub t} the true mass fraction of steam, and all units are metric but the heat flux is in MWm{sup -2}. A simple kinetic model has been developed to estimate X{sub t} as a function of G, X, X{sub i}, and X{sub O}, where X{sub i} is the inlet quality and X{sub O} represents the quality at the Onset of Significant Vaporization (OSV) which is estimated from the Saha-Zuber (S-Z) correlation. The model is based on a rate equation for vaporization suggested by, and consistent with, the S-Z correlation and contains no adjustable parameters. When X{sub i}X{sub O}, X{sub t} depends on X{sub i}, a nonlocal variable, and, in this case, CHF, although determined by local conditions, obeys a nonlocal correlation. This model appears to be satisfactory for pressures less than 110 bar, where the S-Z correlation is known to be reliable. Above 110 bar the method of calculating X{sub O}, and consequently X{sub t}, appears to fail, so this approach can not be applied to high pressure CHF data. Above 35 bar, the bulk of the available data lies in the high CHF regime while, at pressures less than 35 bar, almost all of the available data lie in the low CHF regime and appear to be nonlocal.

Critical heat flux of forced convection boiling in an oscilating acceleration field. Pt. 1

International Nuclear Information System (INIS)

Otsuji, T.; Kurosawa, A.

1982-01-01

The influence of periodically varying acceleration on critical heat flux (CHF) of Freon-113 flowing upward in a uniformly heated vertical annular channel has been studied experimentally. The freon loop was oscillated vertically to determine the ratio of CHF in the oscillating acceleration field to the corresponding stationary value. The amplitude of inlet flow oscillation induced by variation of acceleration, which causes early CHF, is proportional to the acceleration amplitude. The dependence of inlet flow rate on the oscillating acceleration decreases with increasing inlet subcooling, and no oscillation of inlet flow is observed in the case of negative exit quality (subcooled boiling). Nevertheless the degradation of CHF is more remarkable in the low quality region. This result suggests the necessity to introduce an other mechanism of early CHF than flow oscillation. (orig.)

Review of issues relating to enhancement of critical heat flux in nuclear reactors - 15220

International Nuclear Information System (INIS)

Hasan, S.M.K.; Chowdhury, M.A.Z.; Sarkar, M.A.R.

2015-01-01

Critical heat flux (CHF) is manifested by sharp reduction in heat transfer coefficient and a drastic rise in fuel rod surface temperature, which may lead to failure of the fuel cladding and is of immense importance in reactor design and operation. In this paper, the existing literature on CHF is studied and analyzed. Different modeling techniques such as geometry modeling and fluid modeling are highlighted. CHF enhancement techniques such as swirl flow, porous coating, nano-fluids or mechanical vibrations are analyzed and their suitability for different applications is discussed. The advances in prediction methods for tubes and rod bundles are also presented. CHF correlations and their applications have been compiled to estimate the CHF for fluids other than water or other geometries. Finally, future research needs are identified and presented. (authors)

Critical heat flux and flow pattern for water flow in annular geometry

International Nuclear Information System (INIS)

Park, Jae Wook; Baek, Won Pil; Chang, Soon Heung

1996-01-01

An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for upward flow

A formal approach for the prediction of the critical heat flux in subcooled water

Energy Technology Data Exchange (ETDEWEB)

Lombardi, C. [Polytechnic of Milan (Italy)

1995-09-01

The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon.

A formal approach for the prediction of the critical heat flux in subcooled water

International Nuclear Information System (INIS)

Lombardi, C.

1995-01-01

The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon

Critical heat flux of subcooled flow boiling in narrow rectangular channels

International Nuclear Information System (INIS)

Kureta, Masatoshi; Akimoto, Hajime

1999-01-01

In relation to the high-heat-load devices such as a solid-target cooling channel of a high-intensity neutron source, burnout experiments were performed to obtain critical heat flux (CHF) data systematically for vertical upward flow in one-side heated rectangular channels. One of the objectives of this study was to study an extensibility of existing CHF correlations and models, which were proposed for a round tube, to rectangular channels for design calculation. Existing correlations and models were reviewed and compared with obtained data. Sudo's thin liquid layer dryout model, Griffel correlation and Bernath correlation were in good agreement with the experimental data for short-heated-length and low inlet water temperature conditions. (author)

Flow regimes and mechanistic modeling of critical heat flux under subcooled flow boiling conditions

Science.gov (United States)

Le Corre, Jean-Marie

Thermal performance of heat flux controlled boiling heat exchangers are usually limited by the Critical Heat Flux (CHF) above which the heat transfer degrades quickly, possibly leading to heater overheating and destruction. In an effort to better understand the phenomena, a literature review of CHF experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available data. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. It is postulated that a high local wall superheat occurs locally in a dry area of the heated wall, due to a cyclical event inherent to the considered CHF two-phase flow regime, preventing rewetting (Leidenfrost effect). The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow. A numerical model using a two-dimensional transient thermal analysis of the heater undergoing nucleation was developed to mechanistically predict CHF in the case of a bubbly flow regime. In this type of CHF two-phase flow regime, the high local wall superheat occurs underneath a nucleating bubble at the time of bubble departure. The model simulates the spatial and temporal heater temperature variations during nucleation at the wall, accounting for the stochastic nature of the boiling phenomena. The model has also the potential to evaluate

Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

Energy Technology Data Exchange (ETDEWEB)

Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

2013-10-15

Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

High heat flux tests at divertor relevant conditions on water-cooled swirl tube targets

International Nuclear Information System (INIS)

Schlosser, J.; Boscary, J.

1994-01-01

High heat flux experiments were performed to provide a technology for heat flux removal under NET/ITER relevant conditions. The water-cooled rectangular test sections were made of hardened copper with a stainless steel twisted tape installed inside a circular channel and one-side heated. The tests aimed to investigate the heat transfer and the critical heat flux in the subcooled boiling regime. A CHF data base of 63 values was established. Test results have shown the thermalhydraulic ability of swirl tubes to sustain an incident heat flux up to a 30 MW.m -2 range. (author) 10 refs.; 7 figs

Post-CHF low-void heat transfer of water: measurements in the complete transition boiling region at atmospheric pressure

International Nuclear Information System (INIS)

Johannsen, K.; Meinen, W.

1984-01-01

An experimental investigation of low-void heat transfer of water has been performed in the range of CHF and the minimum stable film boiling temperature. The heat transfer system used consists of a vertically mounted copper tube of 1 cm I.D. and 5 cm length with surface-temperature controlled, indirect Joule heating. Results are presented for upflowing water at inverted annular flow conditions in the inlet subcooling range of 2.5 - 40 0 C and mass flux range of 137-600 kg/m 2 s in terms of boiling curves and heat transfer coefficients versus wall temperature. Heat transfer in the stationary rewetting front, which occurs within the test section during operation in the transition boiling mode, is also dealt with. At high mass flux, occurrence of an inverse rewetting front has been observed. It is also noted that, at fixed location, minimum heat flux observed is usually not associated with the minimum stable film boiling temperature

Boiling Heat Transfer Coefficients of Nanofluids Containing Carbon Nanotubes up to Critical Heat Fluxes

International Nuclear Information System (INIS)

Park, Ki Jung; Lee, Yohan; Jung, Dong Soo; Shim, Sang Eun

2011-01-01

In this study, the nucleate pool boiling heat transfer coefficients (HTCs) and critical heat flux (CHF) for a smooth and square flat heater in a pool of pure water with and without carbon nanotubes (CNTs) dispersed at 60 .deg. C were measured. Tested aqueous nanofluids were prepared using CNTs with volume concentrations of 0.0001%, 0.001%, and 0.01%. The CNTs were dispersed by chemically treating them with an acid in the absence of any polymers. The results showed that the pool boiling HTCs of the nanofluids are higher than those of pure water in the entire nucleate boiling regime. The acid-treated CNTs led to the deposition of a small amount of CNTs on the surface, and the CNTs themselves acted as heat-transfer-enhancing particles, owing to their very high thermal conductivity. There was a significant increase in the CHF- up to 150%-when compared to that of pure water containing CNTs with a volume concentration of 0.001%. This is attributed to the change in surface characteristics due to the deposition of a very thin layer of CNTs on the surface. This layer delays nucleate boiling and causes a reduction in the size of the large vapor canopy around the CHF. This results in a significant increase in the CHF

Review of the critical heat flux correlations for liquid metals

International Nuclear Information System (INIS)

Lee, Yong Bum; Han, H. D.; Chang, W. P.; Kwon, Y. M.

1999-09-01

The CHF phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal reactor (LMR). Especially in the LWR application, many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed over the past three decades. Most of the existing CHF correlations have been developed for light water reactor core applications. Compared with water, liquid metals show a divergent picture of boiling pattern. This can be attributed to the consequence that special CHF conditions obtained from investigations with water cannot be applied to liquid metals. Numerous liquid metal boiling heat transfer and two-phase flow studies have put emphasis on development of models and understanding of the mechanism for improving the CHF predictions. Thus far, no overall analytical solution method has been obtained and the reliable prediction method has remained empirical. The principal objectives of the present report are to review the state of the art in connection with liquid metal critical heat flux under low pressure and low flow conditions and to discuss the basic mechanisms. (author)

Pool boiling characteristics and critical heat flux mechanisms of microporous surfaces and enhancement through structural modification

Science.gov (United States)

Ha, Minseok; Graham, Samuel

2017-08-01

Experimental studies have shown that microporous surfaces induce one of the highest enhancements in critical heat flux (CHF) during pool boiling. However, microporous surfaces may also induce a very large surface superheat (>100 °C) which is not desirable for applications such as microelectronics cooling. While the understanding of the CHF mechanism is the key to enhancing boiling heat transfer, a comprehensive understanding is not yet available. So far, three different theories for the CHF of microporous surfaces have been suggested: viscous-capillary model, hydrodynamic instability model, and dryout of the porous coatings. In general, all three theories account for some aspects of boiling phenomena. In this study, the theories are examined through their correlations with experimental data on microporous surfaces during pool boiling using deionized (DI) water. It was found that the modulation of the vapor-jet through the pore network enables a higher CHF than that of a flat surface based on the hydrodynamic instability theory. In addition, it was found that as the heat flux increases, a vapor layer grows in the porous coatings described by a simple thermal resistance model which is responsible for the large surface superheat. Once the vapor layer grows to fill the microporous structure, transition to film boiling occurs and CHF is reached. By disrupting the formation of this vapor layer through the fabrication of channels to allow vapor escape, an enhancement in the CHF and heat transfer coefficient was observed, allowing CHF greater than 3500 kW/m2 at a superheat less than 50 °C.

An improved mechanistic critical heat flux model for subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

An improved mechanistic critical heat flux model for subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

The effect of the advanced drift-flux model of ASSERT-PV on critical heat flux, flow and void distributions in CANDU bundle subchannels

International Nuclear Information System (INIS)

Hammouda, N.; Rao, Y.F.

2017-01-01

Highlights: • Presentation of the “advanced” drift-flux model of the subchannel code ASSERT-PV. • Study the effect of the drift-flux model of ASSERT on CHF and flow distribution. • Quantify model component effects with flow, quality and dryout power measurements. - Abstract: This paper studies the effect of the drift flux model of the subchannel code ASSERT-PV on critical heat flux (CHF), void fraction and flow distribution across fuel bundles. Numerical experiments and comparison against measurements were performed to examine the trends and relative behaviour of the different components of the model under various flow conditions. The drift flux model of ASSERT-PV is composed of three components: (a) the lateral component or diversion cross-flow, caused by pressure difference between connected subchannels, (b) the turbulent diffusion component or the turbulent mixing through gaps of subchannels, caused by instantaneous turbulent fluctuations or flow oscillations, and (c) the void drift component that occurs due to the two-phase tendency toward a preferred distribution. This study shows that the drift flux model has a significant impact on CHF, void fraction and flow distribution predictions. The lateral component of the drift flux model has a stronger effect on CHF predictions than the axial component, especially for horizontal flow. Predictions of CHF, void fraction and flow distributions are most sensitive to the turbulent diffusion component of the model, followed by the void drift component. Buoyancy drift can be significant, but it does not have as much influence on CHF and flow distribution as the turbulent diffusion and void drift.

Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

International Nuclear Information System (INIS)

Hassan, Y.A.

1987-01-01

To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

On the look-up tables for the critical heat flux in tubes (history and problems)

Energy Technology Data Exchange (ETDEWEB)

Kirillov, P.L.; Smogalev, I.P. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

1995-09-01

The complication of critical heat flux (CHF) problem for boiling in channels is caused by the large number of variable factors and the variety of two-phase flows. The existence of several hundreds of correlations for the prediction of CHF demonstrates the unsatisfactory state of this problem. The phenomenological CHF models can provide only the qualitative predictions of CHF primarily in annular-dispersed flow. The CHF look-up tables covered the results of numerous experiments received more recognition in the last 15 years. These tables are based on the statistical averaging of CHF values for each range of pressure, mass flux and quality. The CHF values for regions, where no experimental data is available, are obtained by extrapolation. The correction of these tables to account for the diameter effect is a complicated problem. There are ranges of conditions where the simple correlations cannot produce the reliable results. Therefore, diameter effect on CHF needs additional study. The modification of look-up table data for CHF in tubes to predict CHF in rod bundles must include a method which to take into account the nonuniformity of quality in a rod bundle cross section.

On the look-up tables for the critical heat flux in tubes (history and problems)

International Nuclear Information System (INIS)

Kirillov, P.L.; Smogalev, I.P.

1995-01-01

The complication of critical heat flux (CHF) problem for boiling in channels is caused by the large number of variable factors and the variety of two-phase flows. The existence of several hundreds of correlations for the prediction of CHF demonstrates the unsatisfactory state of this problem. The phenomenological CHF models can provide only the qualitative predictions of CHF primarily in annular-dispersed flow. The CHF look-up tables covered the results of numerous experiments received more recognition in the last 15 years. These tables are based on the statistical averaging of CHF values for each range of pressure, mass flux and quality. The CHF values for regions, where no experimental data is available, are obtained by extrapolation. The correction of these tables to account for the diameter effect is a complicated problem. There are ranges of conditions where the simple correlations cannot produce the reliable results. Therefore, diameter effect on CHF needs additional study. The modification of look-up table data for CHF in tubes to predict CHF in rod bundles must include a method which to take into account the nonuniformity of quality in a rod bundle cross section

CHF prediction in rod bundles using round tube data

Energy Technology Data Exchange (ETDEWEB)

Souza, Wallen F.; Veloso, Maria A.F.; Pereira, Cláubia; Costa, Antonella L., E-mail: wallenfds@yahoo.com.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat fluxes in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table could be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. (author)

Effects of Al{sub 2}O{sub 3} nanoparticles deposition on critical heat flux of R-123 in flow boiling heat transfer

Energy Technology Data Exchange (ETDEWEB)

Seo, Seok Bin; Bang, In Cheol [School of Mechanical and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST), Ulsan (Korea, Republic of)

2015-06-15

In this study, R-123 flow boiling experiments were carried out to investigate the effects of nanoparticle deposition on heater surfaces on flow critical heat flux (CHF) and boiling heat transfer. It is known that CHF enhancement by nanoparticles results from porous structures that are very similar to layers of Chalk River unidentified deposit formed on nuclear fuel rod surfaces during the reactor operation period. Although previous studies have investigated the surface effects through surface modifications, most studies are limited to pool boiling conditions, and therefore, the effects of porous surfaces on flow boiling heat transfer are still unclear. In addition, there have been only few reports on suppression of wetting for decoupled approaches of reasoning. In this study, bare and Al{sub 2}O{sub 3} nanoparticle-coated surfaces were prepared for the study experiments. The CHF of each surface was measured with different mass fluxes of 1,600 kg/m{sup 2}s, 1,800 kg/m{sup 2}s, 2,100 kg/m{sup 2}s, 2,400 kg/m{sup 2}s, and 2,600 kg/m{sup 2}s. The nanoparticle-coated tube showed CHF enhancement up to 17% at a mass flux of 2,400 kg/m{sup 2}s compared with the bare tube. The factors for CHF enhancement are related to the enhanced rewetting process derived from capillary action through porous structures built-up by nanoparticles while suppressing relative wettability effects between two sample surfaces as a highly wettable R-123 refrigerant was used as a working fluid.

Investigation of Body Force Effects on Flow Boiling Critical Heat Flux

Science.gov (United States)

Zhang, Hui; Mudawar, Issam; Hasan, Mohammad M.

2002-01-01

The bubble coalescence and interfacial instabilities that are important to modeling critical heat flux (CHF) in reduced-gravity systems can be sensitive to even minute body forces. Understanding these complex phenomena is vital to the design and safe implementation of two-phase thermal management loops proposed for space and planetary-based thermal systems. While reduced gravity conditions cannot be accurately simulated in 1g ground-based experiments, such experiments can help isolate the effects of the various forces (body force, surface tension force and inertia) which influence flow boiling CHF. In this project, the effects of the component of body force perpendicular to a heated wall were examined by conducting 1g flow boiling experiments at different orientations. FC-72 liquid was boiled along one wall of a transparent rectangular flow channel that permitted photographic study of the vapor-liquid interface at conditions approaching CHF. High-speed video imaging was employed to capture dominant CHF mechanisms. Six different CHF regimes were identified: Wavy Vapor Layer, Pool Boiling, Stratification, Vapor Counterflow, Vapor Stagnation, and Separated Concurrent Vapor Flow. CHF showed great sensitivity to orientation for flow velocities below 0.2 m/s, where very small CHF values where measured, especially with downflow and downward-facing heated wall orientations. High flow velocities dampened the effects of orientation considerably. Figure I shows representative images for the different CHF regimes. The Wavy Vapor Layer regime was dominant for all high velocities and most orientations, while all other regimes were encountered at low velocities, in the downflow and/or downward-facing heated wall orientations. The Interfacial Lift-off model was modified to predict the effects of orientation on CHF for the dominant Wavy Vapor Layer regime. The photographic study captured a fairly continuous wavy vapor layer travelling along the heated wall while permitting liquid

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

International Nuclear Information System (INIS)

Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

2005-01-01

The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m 2 .s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m 2 .s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux ( 2 .s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water

Study on critical heat flux based on wavelet transform in rectangular narrow channels

International Nuclear Information System (INIS)

Zhou Tao; Ju Zhongyun; Zhang Lei; Li Jingjing; Sheng Cheng; Xiao Zejun

2014-01-01

Critical heat flux is very important for nuclear reactor safety, and observing temperature rise rate is a feasible method. Through using the wavelet transform to analyze the CHF temperature rise curves in rectangular narrow channels, it can remove relative weaker interference and effectively judge CHF. Rectangular narrow channel can strengthen heat transfer and reduce CHF, whose characteristics are proved by, temperature rise curves analyzed by wavelet transform. Respectively applying Daubechies function and Haar function is for guarantee the accuracy of the wavelet analysis, and Daubechies function is more accurate than Haar function in the detail signal processing from results. While the wavelet analysis and experimental results are compared and found in good agreement with the experimental results. (authors)

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

2017-05-15

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Experimental study of flow instability and CHF in a natural circulation system with subcooled boiling

International Nuclear Information System (INIS)

Yang, R.C.; Shi, D.Q.; Lu, Z.Q.; Zheng, R.C.; Wang, Y.

1996-01-01

Experimental study has been performed to investigate flow instability and critical heat flux (CHF) in a natural circulation system with subcooled boiling. In the experiments three kinds of heated sections were used. Freon-12 was used as the working medium. The experiments show which one of the two phenomena, flow instability and CHF condition, may first occur in the system depends on not only the heat input power to the heated section and the parameters of the working medium, but also the construction of the heated section. The occurrence of the flow instability mainly depends on the total heat input power to the heated section and the CHF condition is mainly caused by the local heat flux of the heated section. In the experiments two kinds of flow instability, flow instability with high frequency and flow instability with low frequency, were found. But they all belong to density wave instability. The influence of the parameters of the working medium on the onset of the flow instability and CHF condition in the system were investigated. The stability boundaries were determined through the experiments. By means of dimensional analysis of integral equations, a common correlation describing the threshold condition of onset of the flow instability was obtained

Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

1999-10-01

In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

Integrating artificial neural networks and empirical correlations for the prediction of water-subcooled critical heat flux

International Nuclear Information System (INIS)

Mazzola, A.

1997-01-01

The critical heat flux (CHF) is an important parameter for the design of nuclear reactors, heat exchangers and other boiling heat transfer units. Recently, the CHF in water-subcooled flow boiling at high mass flux and subcooling has been thoroughly studied in relation to the cooling of high-heat-flux components in thermonuclear fusion reactors. Due to the specific thermal-hydraulic situation, very few of the existing correlations, originally developed for operating conditions typical of pressurized water reactors, are able to provide consistent predictions of water-subcooled-flow-boiling CHF at high heat fluxes. Therefore, alternative predicting techniques are being investigated. Among these, artificial neural networks (ANN) have the advantage of not requiring a formal model structure to fit the experimental data; however, their main drawbacks are the loss of model transparency ('black-box' character) and the lack of any indicator for evaluating accuracy and reliability of the ANN answer when 'never-seen' patterns are presented. In the present work, the prediction of CHF is approached by a hybrid system which couples a heuristic correlation with a neural network. The ANN role is to predict a datum-dependent parameter required by the analytical correlation; ; this parameter was instead set to a constant value obtained by usual best-fitting techniques when a pure analytical approach was adopted. Upper and lower boundaries can be possibly assigned to the parameter value, thus avoiding the case of unexpected and unpredictable answer failure. The present approach maintains the advantage of the analytical model analysis, and it partially overcomes the 'black-box' character typical of the straight application of ANNs because the neural network role is limited to the correlation tuning. The proposed methodology allows us to achieve accurate results and it is likely to be suitable for thermal-hydraulic and heat transfer data processing. (author)

Critical heat flux with subcooled boiling of water at low pressure

International Nuclear Information System (INIS)

Chen Yuzhou; Zhou Runbin; Hao Laomi; Chen Haiyan

1997-01-01

The critical heat flux experiment has been performed in round tubes of 10 and 16 mm in diameter with different heating length, covering the range of pressure 1.5-16.7 bar, velocity 1.4-15.4 m/s and exit subcooling 30-136 K. The experimental data and empirical correlations are presented. Based on the results an evaluation of some correlations and 1995 CHF look-up table is made. For the conditions tested the effect of diameter on the critical heat flux is found to be related to the liquid velocity. (author)

Influence of the Particle Length of Carbon Nanotube for Pool Boiling Critical Heat Flux Enhancement of Nanofluids

International Nuclear Information System (INIS)

Park, Sung Seek; Kim, Yong Hwan; Kim, Nam Jin

2013-01-01

The results of this experiment were that the CHF of the two nanofluids increased along with the volumetric fraction until 0.001 vol%, and the two types of nanofluids are the highest CHF at 0.001 vol%. Also, the results show clearly that the rate of CHF increase of the CM-100 MWCNT nanofluid with longer-length nanoparticles is higher than that of the CM-95 MWNCT nanofluid. These results indicate that the length of carbon nanotube influences the pool boiling CHF of carbon nanotube nanofluid and that long-length MWCNT, as above-noted, offers a superior effect in this regard. Boiling heat transfer is used in a variety of industrial processes and applications, such as refrigeration, power generation, heat exchangers, cooling of high-power electronics components and cooling of nuclear reactors. The critical heat flux (CHF) phenomenon is the thermal limit during a boiling heat transfer phase change; at the CHF point the heat transfer is maximised, followed by a drastic degradation after the CHF point. The consequence is a substantial increase in wall temperature which may result in physical failure phenomenon of heat transfer systems. Therefore, the CHF is important being considered in the cooling device design, such as nuclear reactor and nuclear fuels, steam generators, high-density electronic component, etc. And, CHF enhancement is essential for safety of heat transfer system. Recently, CHF reported increased when applied to the nanofluids, with its high (higher-than-base-fluid) thermal characteristic in the nuclear power plant system. Therefore, in this study, carried out the pool boiling CHF experiments by the particle length using carbon nanotube nanofluids, and the results are compared and analyzed for the CHF enhancement. The pool boiling CHF of experiments of carbon nanotube nanofluids carried out by the length of particles and the various concentrations

A theoretical prediction of critical heat flux in saturated pool boiling during power transients

International Nuclear Information System (INIS)

Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

1987-01-01

Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions is of fundamental interest in the design, operation, and safety of boiling and two-phase flow devices. Presented within this paper are the results of a comprehensive theoretical study specifically conducted to model transient CHF behavior in saturated pool boiling. Thermal energy conduction within a heating element and its influence on the CHF are also discussed. The resultant theory provides new insight into the basic physics of the CHF phenomenon and indicates favorable agreement with the experimental data from cylindrical heaters with small radii. However, the flat-ribbon heater data compared poorly with the present theory, although the general trend was predicted. Finally, various factors that affect the discrepency between the data and the theory are listed

Investigation of Critical Heat Flux in Reduced Gravity Using Photomicrographic Techniques

Science.gov (United States)

Mudawar, Issam; Zhang, Hui

2003-01-01

Experiments were performed to examine the effects of body force on flow boiling critical heat flux (CHF). FC-72 was boiled along one wall of a transparent rectangular flow channel that permitted photographic study of the vapor-liquid interface just prior to CHF. High-speed video imaging techniques were used to identify dominant CHF mechanisms corresponding to different flow orientations and liquid velocities. Six different CHF regimes were identified: Wavy Vapor Layer, Pool Boiling, Stratification, Vapor Counterflow, Vapor Stagnation, and Separated Concurrent Vapor Flow. CHF showed significant sensitivity to orientation for flow velocities below 0.2 m/s, where extremely low CHF values where measured, especially with downward-facing heated wall and downflow orientations. High flow velocities dampened the effects of orientation considerably. The CHF data were used to assess the suitability of previous CHF models and correlations. It is shown the Interfacial Lift-off Model is very effective at predicting CHF for high velocities at all orientations. The flooding limit, on the other hand, is useful at estimating CHF at low velocities and for downflow orientations. A new method consisting of three dimensionless criteria is developed for determining the minimum flow velocity required to overcome body force effects on near-saturated flow boiling CHF. Vertical upflow boiling experiments were performed in pursuit of identifying the trigger mechanism for subcooled flow boiling CHF. While virtually all prior studies on flow boiling CHF concern the prediction or measurement of conditions that lead to CHF, this study was focused on events that take place during the CHF transient. High-speed video imaging and photomicrographic techniques were used to record the transient behavior of interfacial features from the last steady-state power level before CHF until the moment of power cut-off following CHF. The video records show the development of a wavy vapor layer which propagates

Critical heat flux and flow pattern for water flow in annular geometry

International Nuclear Information System (INIS)

Park, J.-W.; Baek, W.-P.; Chang, S.H.

1997-01-01

An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for the upward flow. In addition to the experiment, selected CHF correlations for annuli are assessed based on 1156 experimental data from various sources. The Doerffer et al. (1994); Barnett (1966); Jannsen and Kervinen (1963); Levitan and Lantsman (1977) correlations show reasonable predictions for wide parameter ranges, among which the Doerffer et al. (1994) correlation shows the widest parameter ranges and a possibility of further improvement. However, there is no correlation predicting the low-pressure, low-flow CHF satisfactorily. (orig.)

Evaluated experimental database on critical heat flux in WWER FA models

International Nuclear Information System (INIS)

Artamonov, S.; Sergeev, V.; Volkov, S.

2015-01-01

The paper presents the description of the evaluated experimental database on critical heat flux in WWER FA models of new designs. This database was developed on the basis of the experimental data obtained in the years of 2009-2012. In the course of its development, the database was reviewed in terms of completeness of the information about the experiments and its compliance with the requirements of Rostekhnadzor regulatory documents. The description of the experimental FA model characteristics and experimental conditions was specified. Besides, the experimental data were statistically processed with the aim to reject incorrect ones and the sets of experimental data on critical heat fluxes (CHF) were compared for different FA models. As a result, for the fi rst time, the evaluated database on CHF in FA models of new designs was developed, that was complemented with analysis functions, and its main purpose is to be used in the process of development, verification and upgrading of calculation techniques. The developed database incorporates the data of 4183 experimental conditions obtained in 53 WWER FA models of various designs. Keywords: WWER reactor, fuel assembly, CHF, evaluated experimental data, database, statistical analysis. (author)

Experimental investigations of the post-CHF heat transfer of R-134a flow-boiling in an annulus with spacer grids

International Nuclear Information System (INIS)

Lee, Kwi Lim; Chang, Soon Heung

2009-01-01

An experimental study was performed in the post-CHF condition using R-134a to investigate the effect of spacer grids on post-CHF heat transfer in an annulus channel. The experiments were conducted under the outlet pressures of 1.1 - 2.0 MPa, the mass fluxes of 100 - 400 kg/m 2 s and the inlet temperatures of 25 - 51degC. About 300 data of post-CHF data were obtained in the annular geometry without spacer grids and compared with several post-CHF correlations. The results showed the large prediction uncertainty mainly caused by the cold wall effect, so the empirical correlation for an annulus geometry without spacer grids was developed with the present experimental results. The heat transfer coefficient was calculated based on the heater rod temperature and the saturated vapor property. The average and root-mean-square(RMS) errors of the predictions were 0.17 % and 3.4 %, respectively. The experiments related to the spacer grid effects were performed with an I-type spacer grid and split-swirl mixing vane (with blockage-area ratios of 4.0 and 5.8 %). The spacer grid and mixing vane test results showed the enhancing effect on the heat transfer at the downstream location of the spacers. The experimental results from the spilt-swirl-type grid tests were more effective than the I-type grid tests. This was attributed to enhance the turbulence and increase the heat transfer caused by the mixing vane. (author)

A 3-D CFD approach to the mechanistic prediction of forced convective critical heat flux at low quality

International Nuclear Information System (INIS)

Jean-Marie Le Corre; Cristina H Amon; Shi-Chune Yao

2005-01-01

Full text of publication follows: The prediction of the Critical Heat Flux (CHF) in a heat flux controlled boiling heat exchanger is important to assess the maximal thermal capability of the system. In the case of a nuclear reactor, CHF margin gain (using improved mixing vane grid design, for instance) can allow power up-rate and enhanced operating flexibility. In general, current nuclear core design procedures use quasi-1D approach to model the coolant thermal-hydraulic conditions within the fuel bundles coupled with fully empirical CHF prediction methods. In addition, several CHF mechanistic models have been developed in the past and coupled with 1D and quasi-1D thermal-hydraulic codes. These mechanistic models have demonstrated reasonable CHF prediction characteristics and, more remarkably, correct parametric trends over wide range of fluid conditions. However, since the phenomena leading to CHF are localized near the heater, models are needed to relate local quantities of interest to area-averaged quantities. As a consequence, large CHF prediction uncertainties may be introduced and 3D fluid characteristics (such as swirling flow) cannot be accounted properly. Therefore, a fully mechanistic approach to CHF prediction is, in general, not possible using the current approach. The development of CHF-enhanced fuel assembly designs requires the use of more advanced 3D coolant properties computations coupled with a CHF mechanistic modeling. In the present work, the commercial CFD code CFX-5 is used to compute 3D coolant conditions in a vertical heated tube with upward flow. Several CHF mechanistic models at low quality available in the literature are coupled with the CFD code by developing adequate models between local coolant properties and local parameters of interest to predict CHF. The prediction performances of these models are assessed using CHF databases available in the open literature and the 1995 CHF look-up table. Since CFD can reasonably capture 3D fluid

Study on critical heat flux based on wavelet transform in rectangular narrow channels

International Nuclear Information System (INIS)

Zhou Tao; Ju Zhongyun; Zhang Lei; Li Jingjing; Sheng Cheng; Xiao Zejun

2014-01-01

Critical heat flux is very important for the safety of nuclear reactor, and observing temperature rise rate is a feasible method. The wavelet transform is used to analyze the CHF temperature rise curves in rectangular narrow channels, which can remove relative weaker interference and effectively judge CHF. Rectangular narrow channel can strengthen heat transfer and reduce CHF, whose characteristics are proved by temperature rise curves analyzed by wavelet transform. Respectively applying Daubechies function and Haar function is to guarantee the accuracy of the wavelet analysis, and Daubechies function is more accurate than Haar function in the detail signal processing from results. While the wavelet analysis and experimental results are compared and found in good agreement with the experimental results. (authors)

Experiments on Critical Heat Flux for CAREM -25 Reactor

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

Critical heat flux in a vertical annulus under low upward flow and near atmospheric pressure

International Nuclear Information System (INIS)

Schoesse, T.; Aritomi, Masanori; Lee, Sang-Ryoul; Kataoka, Yoshiaki; Yoshioka, Yuzuru; Chung, Moon-Ki.

1997-01-01

As future boiling water reactors (BWR), concepts of evolutional ABWR (ABWR-IER) and natural circulation BWR (JSBWR) have been investigated in order to reduce their construction cost and simplify their maintenance and inspection procedures. One of the promised features of the design of the evolutional ABWR is to reduce the number of internal pumps and to remove the Motor Generation (MG) sets. These design changes may induce boiling transition in the fuel rods of reactor core during a pump trip transient due to the more rapid flow coastdown characteristics than these of the present design. In addition, the understanding of critical heat flux (CHF) is one important subject to grasp safety margin during the start-up for the natural circulation BWR and to establish the rational start-up procedure in which thermo-hydraulic instabilities can be suppressed. The present study is to clarify CHF characteristics under low velocity conditions. CHF measurements were conducted in a vertical upward annulus channel composed of an inner heated rod and an outer tube made of glass. CHF data were obtained repeatedly under the condition of stable inlet flow to examine statistically their reproducibility. The flow regime was investigated from flow observation and measurement of differential pressure fluctuation. The CHF data are correlated with the flow regime transition. It was clear from the obtained flow pattern and the CHF data that the CHF behavior could be classified into specified regions by the mass flux and inlet subcooling conditions. A CHF correlation was developed and agreed with other researchers' data within acceptable error. (author)

Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle

International Nuclear Information System (INIS)

Onder, E.N.; Leung, L.K.H.

2013-01-01

Boiling-Length-Average (BLA) Critical Heat Flux (CHF) values for the CANFLEX bundle at cross-sectional average subcooled conditions have been evaluated using the ASSERT-PV subchannel code. The predicted BLA CHF values supplement experimental BLA CHF values obtained with full-scale bundle simulators at saturated conditions in developing a BLA CHF correlation applicable over the interested range of cross-sectional average thermodynamic quality in regional overpower protection (ROP) trip and safety analyses. The BLA CHF correlation exhibits similar characteristics to those observed in tubes at subcooled and saturated conditions. Applying this correlation has led to similar prediction accuracy in dryout power to that using the BLA CHF-data-based correlation at saturated conditions. However, it provides improved prediction accuracy in dryout power at dryout conditions near saturation compared to the BLA CHF-data-based correlation (which tends to underpredict the dryout power)

Effects of Liquid Metal Fin on Critical Heat Flux under IVR-ERVC Condition

Energy Technology Data Exchange (ETDEWEB)

Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

The molten fuel is relocated to bottom of reactor vessel after core is damaged and not cooled continuously. In-vessel retention through external reactor vessel cooling (IVR-ERVC) is presented to terminate the progression of accidents by removing the decay heat. IVR-ERVC is suitable for small size reactors like AR-600, AP-1000. There is uncertainty for high power reactor like APR-1400 and CAP-1400. This uncertainty originates from the thermal margin between the CHF value and real heat flux on the reactor vessel under severe accidents. The main mechanism of heat removal on IVR-ERVC strategy is boiling on the outer wall of reactor vessel. The boiling heat transfer is limited due to the CHF phenomenon. There should be an enough margin for preventing the CHF in boiling heat transfer systems. The CHF tests for IVR-ERVC system were conducted to confirm or increase the thermal margin. The design of thermal insulator was changed to vent the vapor smoothly. Forming the coating layer on the vessel surface was proposed to enhance the CHF margin. The liquid metal was designed to flood the space around the reactor vessel. The liquid metal has high boiling point and superb thermal conductivity in comparison with the coolant. In this work, experimental tests were conducted to validate the CFD results about the IVR-ERVC system with liquid metal. The behavior of vapor was observed to predict the tendency of CHF increase with small-scaled facility to simulate the IVR-ERVC system.

Classification and prediction of the critical heat flux using fuzzy theory and artificial neural networks

International Nuclear Information System (INIS)

Moon, Sang Ki; Chang, Soon Heung

1994-01-01

A new method to predict the critical heat flux (CHF) is proposed, based on the fuzzy clustering and artificial neural network. The fuzzy clustering classifies the experimental CHF data into a few data clusters (data groups) according to the data characteristics. After classification of the experimental data, the characteristics of the resulting clusters are discussed with emphasis on the distribution of the experimental conditions and physical mechanism. The CHF data in each group are trained in an artificial neural network to predict the CHF. The artificial neural network adjusts the weight so as to minimize the prediction error within the corresponding cluster. Application of the proposed method to the KAIST CHF data bank shows good prediction capability of the CHF, better than other existing methods. ((orig.))

CHF enhancement in pool boiling of nanofluid : effect of nanoparticle-coating on heating surface

International Nuclear Information System (INIS)

Kim, Hyung Dae; Kim, Moo Hwan

2005-01-01

Recently researches to enhance CHF using the nanofluid, a new kind of heat transfer fluid in which nano-particles are uniformly and stably dispersed, were attempted. You showed that nanofluid, containing only 0.005 g/l of alumina nanoparticle, make the dramatic increase (∼200%) in CHF in pool boiling at the pressure of 2.89 psia (Tsat=60 .deg. C). They concluded that the abnormal CHF enhancement of nanofluid cannot be explained with any existing models of CHF. Vassallo performed the experimental studies on pool boiling heat transfer in water-SiO 2 nanofluid under atmospheric pressure. They showed a remarkable increase in CHF for nanofluid and also found that the stable film boiling at temperatures close to the melting point of the boiling surface are achievable with the nanofluid. After the experiments, they observed that the formation of the thin silica coating on the wire heater was occurred. This paper focuses on the experimental study of the effect of nanoparticle-coating on CHF enhancement in pool boiling of nanofluid. In this regard, pool boiling CHF values are measured and compared (a) from bare heater immersed in nanofluid and (b) from nanoparticle-coated heater, which is generated by deposition of suspended nanoparticles during pool boiling of nanofluid, immersed in pure water, and (c) from nanoparticle-coated heater immersed in nanofluid. And the microstructure of each heating surface is investigated from photography taken using SEM

An experimental study of forced convective flow boiling CHF in nanofluid

International Nuclear Information System (INIS)

Ahn, Hoseon; Kim, Seontae; Jo, Hangjin; Kim, Dongeok; Kang, Soonho; Kim, Moohwan

2008-01-01

Recently the enhancement of CHF (critical heat flux) in nanofluids under the pool boiling condition is known as a result of nanoparticle deposition on the heating surface. The deposition phenomenon of nanoparticles on the heating surface is induced dominantly by the vigorous boiling on the heating surface. Considering the importance of flow boiling conditions in various practical heat transfer applications, an experimental study was performed to verify whether or not the enhancement of CHF in nanofluids exists in a forced convective flow boiling condition. The nanofluid used in this research was Al 2 O 3 -water dispersed by the ultra-sonic vibration method in very low concentration (0.01% Vol). A heater specimen was made of a copper block easily detachable to look into the surface condition after the experiment. The heating method was a thermal-heating made with a conductive material. The flow channel took a rectangular type (10mm x 10mm) and had a length of 1.2 m to assure a hydrodynamically fully-developed region. In result, CHF in the nanofluid under the forced convective flow boiling condition has been enhanced distinctively along with the effect of flow rates. To reason the CHF increase in the nanofluids, the boiling surface was investigated thoroughly with the SEM image. (author)

One-dimensional critical heat flux concerning surface orientation and gap size effects

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Hoon; Suh, Kune Y. E-mail: kysuh@snu.ac.kr

2003-12-01

Tests were conducted to examine the critical heat flux (CHF) on a one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180 deg.) to the vertical position (90 deg.), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle at which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis. This correlation agrees with the experimental data of this study within {+-}20%.

Mechanisms and predictions for subcooled flow boiling CHF

International Nuclear Information System (INIS)

Liu, Wei; Nariai, Hideki; Inasaka, Fujio

2000-01-01

Corresponding to the two kinds of flow pattern reported in literature for subcooled flow boiling, two kinds of CHF triggering mechanism are considered existing with working in different working scope. On the base of a criterion proposed recently by the present authors, subcooled flow boiling data firstly are categorized into two groups by judging whether the first kind or the second kind of flow pattern is established. Possible CHF triggering mechanisms and prediction methods for the two kinds of flow pattern condition are discussed. By considering both the flow pattern development and CHF triggering mechanism, a detailed data categorization is carried out. The corresponding CHF occurrence properties in different data groups are summarized. Parametric trends are reviewed for the first and second kind of data group working condition respectively. Mass flux, pressure, inlet subcooling and inner diameter show almost same effects in the two different working conditions, while the ratio of heated length to diameter's effects on CHF show to be different. Research for the L/D effect on the CHF transverse the interface of the different data groups is carried out. (author)

Study on Enhancement of Sub-Cooled Flow Boiling Heat Transfer and Critical Heat Flux of Solid-Water Two-Phase Mixture

International Nuclear Information System (INIS)

Yasuo Koizumi; Hiroyasu Ohtake; Tomoyuki Suzuki

2002-01-01

The influence of particle introduction into a subcooled water flow on boiling heat transfer and critical heat flux (CHF) was examined. When the water velocity was low, the particles crowded on the bottom wall of the flow channel and flowed just like sliding on the wall. When the water velocity was high, the particles were well dispersed in the water flow. In the non-boiling region, the heat transfer was augmented by the introduction of the particles into the water flow. As the introduction of the particles were increased, the augmentation was also increased in the high water flow rate region. However, it was independent upon the particle introduction rate in the low water flow rate region. The onset of boiling was delayed by the particle inclusion. The boiling heat transfer was enhanced by the particles. However, it was rather decreased in the high heat flux fully-developed-boiling region. The CHF was decreased by the particle inclusion in the low water flow region and was not affected in the high water flow region. (authors)

A study on the critical heat flux for annuli and round tubes under low pressure conditions

International Nuclear Information System (INIS)

Park, Jae Wook

1997-02-01

This study aims to reveal the characteristics of the critical heat flux (CHF) of internally heated concentric annuli and vertical round tubes in low-pressure and low-flow (LPLF) conditions. Although many efforts have been devote to the subject of the CHF during the last forty years, the information on the CHF phenomenon for LPLF conditions is still very limited. The applicable ranges of the CHF correlations for annuli and round tubes are concentrate on the operating conditions of nuclear power plant (NPP), namely high-pressure and high-flow (HPHF) conditions. these facts promoted to collect the reliable CHF data for LPLF conditions for both annuli and round tubes. The critical heat flux data for vertical flow boiling of water in annuli and round tubes at low pressures and low mass fluxes show the following trends: The observed CHF mechanism for annuli was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. The observed parametric trends for annuli are consistent with the previous understanding except that the CHF for downward flow is considerably lower (up to 40%) than that for upward flow. The critical quality is much lower than that for round tubes at the same inlet conditions. The observed parametric trends for round tubes are generally consistent with the previous understanding except for system pressure an tube diameter effect. For the system pressure effect, it is observed that the pressure effect is complicated but not so large, whereas the existing CHF correlations do not present the parametric trend exactly. For tube diameter effect, the decreasing trends of CHF with respect to tube diameter was the general understanding so far, but in this region the CHF show a increasing trend of tube diameter. The prediction and the parametric trend analyses are performed by two view points, I.e., for fixed inlet conditions and for local

Enhanced CHF with Bubble Cutter and Artificial Flow in Nuclear Plants

International Nuclear Information System (INIS)

Jung, Chan Hee; Suh, Kune Y.

2013-01-01

The main goal of this paper is to body out the notions of forced convection system for enhanced local streams and air bubbles cutting (and/or pushing, breaking) system to explain how CHF can be improved and how those bubble cutter systems are applicable to NPPs. In this paper, the bubble cutter system and an artificial flow system which can cut (and/or push and break) air bubbles is bodied out to drag bubbles. It also make the surface wet condition of heated surfaces and improve heat transfer and prevent on creation of bubbles on the heated surfaces or heat exchangers or reactor cores. Namely, concepts and application methods to increase CHF are presented for NPPs. Enhanced critical heat flux (CHF) is one of our prospective aims for nuclear power plants (NPPs). Previous work has studied the flow boiling CHF enhancement with surfactant solutions under atmospheric pressure because surfactant solutions or surface conditions have an effect on the behavior of occurrence air bubbles on a heated surface. Another possible improvement is to improve efficiency of heat transfer or to body out some types of bubble breaking (and/or pushing, breaking) systems or an artificial flow of fluid that can tear off air bubbles or push hot liquid and bubbles on a surface of heater. During this study, it will be observed that those possible structures can elicit increased CHF by means of maintenance of contact with a coolant such as water

Development of classification and prediction methods of critical heat flux using fuzzy theory and artificial neural networks

International Nuclear Information System (INIS)

Moon, Sang Ki

1995-02-01

This thesis applies new information techniques, artificial neural networks, (ANNs) and fuzzy theory, to the investigation of the critical heat flux (CHF) phenomenon for water flow in vertical round tubes. The work performed are (a) classification and prediction of CHF based on fuzzy clustering and ANN, (b) prediction and parametric trends analysis of CHF using ANN with the introduction of dimensionless parameters, and (c) detection of CHF occurrence using fuzzy rule and spatiotemporal neural network (STN). Fuzzy clustering and ANN are used for classification and prediction of the CHF using primary system parameters. The fuzzy clustering classifies the experimental CHF data into a few data clusters (data groups) according to the data characteristics. After classification of the experimental data, the characteristics of the resulted clusters are discussed with emphasis on the distribution of the experimental conditions and physical mechanisms. The CHF data in each group are trained in an artificial neural network to predict the CHF. The artificial neural network adjusts the weight so as to minimize the prediction error within the corresponding cluster. Application of the proposed method to the KAIST CHF data bank shows good prediction capability of the CHF, better than other existing methods. Parametric trends of the CHF are analyzed by applying artificial neural networks to a CHF data base for water flow in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. In order to remove the necessity of data classification, Katto and Groeneveld et al.'s dimensionless parameters are introduced in training the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS error of 8.9%, 13.1%, and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local

Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

International Nuclear Information System (INIS)

Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.

2010-01-01

A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].

Analysis of relations for heat transfer at the post-CHF regime

Energy Technology Data Exchange (ETDEWEB)

Dorokhovich, S. L. [Obninsk State Techical Univ., Obninsk (Russian Federation); Kirillov, P. L. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

2003-07-01

Information about heat transfer rates in two-fluid flow at the post-CHF regime is important for analysis of accidents of water-cooled nuclear reactors as well as calculations of steam generators (liquid metal-water). It is complicate to create analytical methods because there is a variety of two-fluid flow regimes dictated by channel size, pressure, mass flow rate, heat flux, droplet spectrum, flow quality, other factors and, finally, by crisis type. At crisis in subcooled liquid or low quality two-phase flows Inverted Annular Film Boiling (IAFB) takes place when liquid flow is separated from the wall by a thin superheated vapor film. For dispersed flow crisis is usually related to drying a liquid film moving along the heated surface (wall). In both cases two-phase flow is thermodynamically nonequilibrium since the temperatures of phases (liquid and vapor) are different. The mean (at the rate of heat content) flow temperature is not the determining parameter. Different boundary conditions of experiment s uniform heat fluxes or 'hot' spots, are able to lead to different relations for the heat transfer coefficient. Last years the great number of semiempirical models were elaborated, that become more and more complicate. It is difficult to examine many parameters of the models. An agreement between final results and separate experimental data is not yet the evidence for the verity of extension while prerequisites taken in the models are often doubtful and hardly examined. Thus the correlations obtained from experimental data, for example, are used in practice. The analysis of relations for heat transfer at the Deteriorated Heat Transfer (DHT) regime, the comparison of relations with the data of look up tables made on the basis of the Institute of Physics and Power Engineering and Chalk River Laboratories experimental data banks are the objective of current report.

Analysis of relations for heat transfer at the post-CHF regime

International Nuclear Information System (INIS)

Dorokhovich, S. L.; Kirillov, P. L.

2003-01-01

Information about heat transfer rates in two-fluid flow at the post-CHF regime is important for analysis of accidents of water-cooled nuclear reactors as well as calculations of steam generators (liquid metal-water). It is complicate to create analytical methods because there is a variety of two-fluid flow regimes dictated by channel size, pressure, mass flow rate, heat flux, droplet spectrum, flow quality, other factors and, finally, by crisis type. At crisis in subcooled liquid or low quality two-phase flows Inverted Annular Film Boiling (IAFB) takes place when liquid flow is separated from the wall by a thin superheated vapor film. For dispersed flow crisis is usually related to drying a liquid film moving along the heated surface (wall). In both cases two-phase flow is thermodynamically nonequilibrium since the temperatures of phases (liquid and vapor) are different. The mean (at the rate of heat content) flow temperature is not the determining parameter. Different boundary conditions of experiment s uniform heat fluxes or 'hot' spots, are able to lead to different relations for the heat transfer coefficient. Last years the great number of semiempirical models were elaborated, that become more and more complicate. It is difficult to examine many parameters of the models. An agreement between final results and separate experimental data is not yet the evidence for the verity of extension while prerequisites taken in the models are often doubtful and hardly examined. Thus the correlations obtained from experimental data, for example, are used in practice. The analysis of relations for heat transfer at the Deteriorated Heat Transfer (DHT) regime, the comparison of relations with the data of look up tables made on the basis of the Institute of Physics and Power Engineering and Chalk River Laboratories experimental data banks are the objective of current report

A scaling law for the local CHF on the external bottom side of a fully submerged reactor vessel

International Nuclear Information System (INIS)

Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

1997-01-01

A scaling law for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water has been developed from the results of an advanced hydrodynamic CHF model for pool boiling on a downward facing curved heating surface. The scaling law accounts for the effects of the size of the vessel, the level of liquid subcooling, the intrinsic properties of the fluid, and the spatial variation of the local critical heat flux along the heating surface. It is found that for vessels with diameters considerably larger than the characteristic size of the vapor masses, the size effect on the local critical heat flux is limited almost entirely to the effect of subcooling associated with the local liquid head. When the subcooling effect is accounted for separately, the local CHF limit is nearly independent of the vessel size. Based upon the scaling law developed in this work, it is possible to merge, within the experimental uncertainties, all the available local CHF data obtained for various vessel sizes under both saturated and subcooled boiling conditions into a single curve. Applications of the scaling law to commercial-size vessels have been made for various system pressures and water levels above the heated vessel. Over the range of conditions explored in this study, the local CHF limit is found to increase by a factor of two or more from the bottom center to the upper edge of the vessel. Meanwhile, the critical heat flux at a given angular position of the heated vessel is also found to increase appreciably with the system pressure and the water level

Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

Energy Technology Data Exchange (ETDEWEB)

Limin, Zheng [Shanghai Nuclear Engineering Research and Design Inst., SH (China); Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime

1997-08-01

The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms{sup -1} and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D{sub hy}=0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm{sup -2}s{sup -1} (water flow velocity, 5-15 ms{sup -1}); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within {+-}18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within {+-}16%. (author)

Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

International Nuclear Information System (INIS)

Zheng Limin; Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime.

1997-08-01

The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms -1 and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D hy =0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm -2 s -1 (water flow velocity, 5-15 ms -1 ); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within ±18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within ±16%. (author)

Effects of Microencapsulated Phase Change Material (MPCM) on Critical Heat Flux in Pool Boiling

International Nuclear Information System (INIS)

Park, Sung Dae; Kim, Seong Man; Kang, Sarah; Lee, Seung Won; Seo, Han; Bang, In Cheol

2011-01-01

Thermal power is limited by critical heat flux (CHF) in the nuclear power plant. And the in-vessel retention by external reactor vessel cooling (IVR-ERVC) is applied in some nuclear power plants; AP600, AP1000, Loviisa and APR1400. The heat removal capacity of IVR-ERVC is also restricted by CHF. So, it is essential to get CHF margin to improve an economics and a safety of the plant. There are some typical approaches to enhance CHF: vibrating the heater or fluid, coating with porous media on the heater surface, applying an electric field. The recent study related to the CHF is focus on using the nanofluid. In this paper, the new approach was investigated by using the microencapsulated phase change material (MPCM). MPCM is the particles whose diameter is from 0.1μm to 1000μm. The MPCM consists of the core material and the shell material. The core material can be solid, liquid, gas or even the mixture. The solid paraffin is the best candidate as the core material due to its stable chemical and thermal properties. And the shell material is generally synthesized polymer of about several micrometers in thickness. The most interesting feature of the MPCM is that the latent heat associated with the solid-liquid phase change is related to the heat transfer. When the MPCM is dispersed into the carrier fluid, a kind of suspension named as microencapsulated phase change slurry (MPCS) is formed. The study on the MPCS was conducted in field of both the heat transfer fluids and energy storage media. It is inspired by the fact that the latent heat can serve distribution to the additional CHF margin. The purpose of this work is to confirm whether or not the CHF is enhanced

A dry-spot model of critical heat flux and transition boiling in pool and subcooled forced convection boiling

International Nuclear Information System (INIS)

Ha, Sang Jun

1998-02-01

A new dry-spot model for critical heat flux (CHF) is proposed. The new concept for dry area formation based on Poisson distribution of active nucleation sites and the critical active site number is introduced. The model is based on the boiling phenomena observed in nucleate boiling such as Poisson distribution of active nucleation sites and formation of dry spots on the heating surface. It is hypothesized that when the number of bubbles surrounding one bubble exceeds a critical number, the surrounding bubbles restrict the feed of liquid to the microlayer under the bubble. Then a dry spot of vapor will form on the heated surface. As the surface temperature is raised, more and more bubbles will have a population of surrounding active sites over the critical number. Consequently, the number of the spots will increase and the size of dry areas will increase due to merger of several dry spots. If this trend continues, the number of effective sites for heat transport through the wall will diminish, and CHF and transition boiling occur. The model is applicable to pool and subcooled forced convection boiling conditions, based on the common mechanism that CHF and transition boiling are caused by the accumulation and coalescences of dry spots. It is shown that CHF and heat flux in transition boiling can be determined without any empirical parameter based on information on the boiling parameters such as active site density and bubble diameter, etc., in nucleate boiling. It is also shown that the present model well represents actual phenomena on CHF and transition boiling and explains the mechanism on how parameters such as flow modes (pool or flow) and surface wettability influence CHF and transition boiling. Validation of the present model for CHF and transition boiling is achieved without any tuning parameter always present in earlier models. It is achieved by comparing the predictions of CHF and heat flux in transition boiling using measured boiling parameters in nucleate

Effect of flow obstacles with various leading and trailing edges on critical heat flux

International Nuclear Information System (INIS)

Pioro, I.L.; Groeneveld, D.C.; Groeneveld, D.C.; Cheng, S.C.; Antoshko, Y.V.

2001-01-01

A joint investigation has been performed by the University of Ottawa and Chalk River Laboratories that examined the effect of the shape of the leading and trailing edges of the turbulence enhancing devices ('flow obstacles') on critical heat flux (CHF). The objective of this study was to gain a better overall understanding of the limit of CHF improvement for various obstacle designs and the impact of flow conditions on the improvements. (author)

Optimizing critical heat flux enhancement through nano-particle-based surface modifications

International Nuclear Information System (INIS)

Truong, B.; Hu, L. W.; Buongiorno, J.

2008-01-01

Colloidal dispersions of nano-particles, also known as nano-fluids, have shown to yield significant Critical Heat Flux (CHF) enhancement. The CHF enhancement mechanism in nano-fluids is due to the buildup of a porous layer of nano-particles upon boiling. Unlike microporous coatings that had been studied extensively, nano-particles have the advantages of forming a thin layer on the substrate with surface roughness ranges from the sub-micron to several microns. By tuning the chemical properties it is possible to coat the nano-particles in colloidal dispersions onto the desired surface, as has been demonstrated in engineering thin film industry. Building on recent work conducted at MIT, this paper illustrates the maximum CHF enhancement that can be achieved based on existing correlations. Optimization of the CHF enhancement by incorporation of key factors, such as the surface wettability and roughness, will also be discussed. (authors)

Application of artificial neural networks in analysis of CHF experimental data in round tubes

International Nuclear Information System (INIS)

Huang Yanping; Chen Bingde; Lang Xuemei; Wang Xiaojun; Shan Jianqiang; Jia Dounan

2004-01-01

Artificial neural networks (ANNs) are applied successfully to analyze the critical heat flux (CHF) experimental data from some round tubes in this paper. A set of software adopting artificial neural network method for predicting CHF in round tube and a set of CHF database are gotten. Comparing with common CHF correlations and CHF look-up table, ANN method has stronger ability of allow-wrong and nice robustness. The CHF predicting software adopting artificial neural network technology can improve the predicting accuracy in a wider parameter range, and is easier to update and to use. The artificial neural network method used in this paper can be applied to some similar physical problems. (authors)

High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

International Nuclear Information System (INIS)

Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

1998-01-01

Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

CHF considerations for highly moderated 100% MOX fuels PWRs

Energy Technology Data Exchange (ETDEWEB)

Saphier, D.; Raymond, P. [CEA Saclay, DMT/SERMA/LETR, Gif-sur-Yvette (France)

1995-09-01

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat flux (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.

Analysis of in-R12 CHF data: influence of hydraulic diameter and heating length; test of Weisman boiling crisis model

International Nuclear Information System (INIS)

Czop, V.; Herer, C.; Souyri, A.; Garnier, J.

1993-09-01

In order to progress on the comprehensive modelling of the boiling crisis phenomenon, Electricite de France (EDF), Commissariat a l'Energie Atomique (CEA) and FRAMATOME have set up experimental programs involving in-R12 tests: the EDF APHRODITE program and the CEA-EDF-FRAMATOME DEBORA program. The first phase in these programs aims to acquire critical heat flux (CHF) data banks, within large thermal-hydraulic parameter ranges, both in cylindrical and annular configurations, and with different hydraulic diameters and heating lengths. Actually, three data banks have been considered in the analysis, all of them concerning in-R12 round tube tests: - the APHRODITE data bank, obtained at EDF with a 13 mn inside diameter, - the DEBORA data bank, obtained at CEA with a 19.2 mm inside diameter, - the KRISTA data bank, obtained at KfK with a 8 mm inside diameter. The analysis was conducted using CHF correlations and with the help of an advanced mathematical tool using pseudo-cubic thin plate type Spline functions. Two conclusions were drawn: -no influence of the heating length on our CHF results, - the influence of the diameter on the CHF cannot be simply expressed by an exponential function of this parameter, as thermal-hydraulic parameters also have an influence. Some calculations with Weisman and Pei theoretical boiling crisis model have been compared to experimental values: fairly good agreement was obtained, but further study must focus on improving the modelling of the influence of pressure and mass velocity. (authors). 12 figs., 4 tabs., 21 refs

Maximum allowable heat flux for a submerged horizontal tube bundle

International Nuclear Information System (INIS)

McEligot, D.M.

1995-01-01

For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose 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

Numerical Analysis on Heat Flux Distribution through the Steel Liner of the Ex-vessel Core Catcher

Energy Technology Data Exchange (ETDEWEB)

Oh, Se Hong; Choi, Choeng Ryul [ELSOLTEC, Yongin (Korea, Republic of); Kim, Byung Jo; Lee, Kyu Bok [KEPCO, Gimcheon (Korea, Republic of); Hwang, Do Hyun [KHNP-CRI, Daejeon (Korea, Republic of)

2016-05-15

In order to prevent material failure of steel container of the core catcher system due to high temperatures, heat flux through the steel liner wall must be kept below the critical heat flux (CHF), and vapor dry-out of the cooling channel must be avoided. In this study, CFD methodology has been developed to simulate the heat flux distribution in the core catcher system, involving following physical phenomena: natural convection in the corium pool, boiling heat transfer and solidification/melting of the corium. A CFD methodology has been developed to simulate the thermal/hydraulic phenomena in the core catcher system, and a numerical analysis has been carried out to estimate the heat flux through the steel liner of the core catcher. High heat flux values are formed at the free surface of the corium pool. However, the heat flux through the steel liner is maintained below the critical heat flux.

Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.

2014-01-01

Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

Two-phase jet impingement cooling for high heat flux wide band-gap devices using multi-scale porous surfaces

International Nuclear Information System (INIS)

Joshi, Shailesh N.; Dede, Ercan M.

2017-01-01

Highlights: • Jet impingement with phase change on multi-scale porous surfaces is investigated. • Porous coated flat, pin-fin, open tunnel, and closed tunnel structures are studied. • Boiling curve, heat transfer coefficient, and pressure drop metrics are reported. • Flow visualization shows vapor removal from the surface is a key aspect of design. • The porous coated pin-fin surface exhibits superior two-phase cooling performance. - Abstract: In the future, wide band-gap (WBG) devices such as silicon carbide and gallium nitride will be widely used in automotive power electronics due to performance advantages over silicon-based devices. The high heat fluxes dissipated by WBG devices pose extreme cooling challenges that demand the use of advanced thermal management technologies such as two-phase cooling. In this light, we describe the performance of a submerged two-phase jet impingement cooler in combination with porous coated heat spreaders and multi-jet orifices. The cooling performance of four different porous coated structures was evaluated using R-245fa as the coolant at sub-cooling of 5 K. The results show that the boiling performance of a pin-fin heat spreader is the highest followed by that for an open tunnel (OPT), closed tunnel (CLT), and flat heat spreader. Furthermore, the flat heat spreader demonstrated the lowest critical heat flux (CHF), while the pin-fin surface sustained a heat flux of 218 W/cm 2 without reaching CHF. The CHF values of the OPT and CLT surfaces were 202 W/cm 2 and 194 W/cm 2 , respectively. The pin-fin heat spreader has the highest two-phase heat transfer coefficient of 97,800 W/m 2 K, while the CLT surface has the lowest heat transfer coefficient of 69,300 W/m 2 K, both at a heat flux of 165 W/cm 2 . The variation of the pressure drop of all surfaces is similar for the entire range of heat fluxes tested. The flat heat spreader exhibited the least pressure drop, 1.73 kPa, while the CLT surface had the highest, 2.17 kPa at a

Testing plan for critical heat flux measurement during in-vessel retention

International Nuclear Information System (INIS)

Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

2015-01-01

In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

The effect of flow direction and magnitude on CHF for low pressure water in thin rectangular channels

International Nuclear Information System (INIS)

Mishima, K.; Nishihara, H.

1985-01-01

Critical heat flow (CHF) at low flow condition can become important in an MTR-type research reactor under a number of accident conditions. Regardless of the initial stages of these accidents, a condition which is basically the decay heat removal by natural convention boiling can develop. Under such conditions, burnout may occur even at a very low heat flow. In view of this, the CHF at low-flow-rate and low-pressure conditions has been studied for water flowing in thin rectangular channels. Experiments were carried out with two types of rectangular test sections, namely, the one heated from one wide side and the other heated from two opposite sides. In order to observe the effects of gravity, CHF was measured both in upflow and downflow. The CHF at complete bottom blockage was also studied. The results indicate that burnout can occur at a much lower heat flux than pool-boiling CHF or than predicted by the conventional correlations. There was observed a minimum CHF at complete bottom blockage and at very low downflow. The low CHF at very low downflow appears to be due to the stagnation of the bubble in the heated section. This fact indicates that special care should be taken in analyzing the boiling phenomenon which occurs when the coolant flow is very low in a low pressure system. (author)

Critical heat flux of R134A and R245FA in a 2.2 mm circular tube

Energy Technology Data Exchange (ETDEWEB)

Tibirica, Cristiano Bigonha; Ribatski, Gherhardt [Universidade de Sao Paulo (EESC/USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Mecanica], E-mails: bigonha@sc.usp.br, ribatski@sc.usp.br; Szczukiewicz, Sylwia; Thome, John Richard [Ecole Polytechnique Federale de Lausanne (LTCM/EPFL) (Switzerland). Lab. of Heat and Mass Transfer], Emails: sylwia.szczukiewicz@epfl.ch, john.thome@epfl.ch

2010-07-01

Critical heat flux (CHF) during flow boiling is generally related to a drastic decrease in the heat transfer coefficient and it is the maximum operational heat flux that can be achieved under safe operation. Due to such a fact, this topic has attracted great attention of the academic society dealing with boiling heat transfer and also in the industrial sector involved with the dissipation of high heat flux densities. In the specific case of high heat flux densities, micro-channel flow boiling is a promising technique for pursuing this objective. The boundary where microscale effects start in flow boiling is still an open issue in the literature and a 3 mm internal diameter (ID) threshold value, as suggested by Kandlikar and Grande (2003) is frequently adopted to characterize this point. Considering the needs for a better understanding of the micro/macro transition, this paper presents new experimental critical heat flux results in saturated flow boiling conditions for a macro/micro-scale tube. The data were obtained in a horizontal 2.20 mm ID stainless steel tube with heating lengths of 361 and 154 mm, R134a and R245fa as working fluids, mass velocities ranging from 100 to 1500 kg/m{sup 2s}, critical heat fluxes from 25 to 300 kW/m2, exit saturation temperatures of 25, 31 and 35 degree C, and critical vapor qualities ranging from 0.55 to 1. The experimental results show that critical heat flux increases with increasing mass velocity and inlet subcooling but decreases with increasing saturation temperature and heated length. The data also indicated a higher CHF for R245fa when compared with R134a at similar conditions. The experimental data were compared against the following CHF predictive methods: Katto and Ohno (1984), Shah (1987), Zhang et al. (2006) and Ong and Thome(2010). Katto and Ohno (1984) and Ong and Thome (2010) best predicted the database with a mean average error smaller than 15%. Both correlations include low and high pressure fluids in their

CHF-KFK-3: A critical heat flux correlation for triangular arrays of rods with tight lattices

International Nuclear Information System (INIS)

Dalle Donne, M.

1991-02-01

High converting PWR's (HCPWR or APWR) are based on fuel elements with rods placed in a tight lattice triangular array. The CHF correlation development previously at KfK for such geometry (CHF-KFK-2 correlation) has been tested against recently performed experiments. The comparison with the Siemens-KWU experiments with rod clusters with spacer grid and six integral spiral ribs supports has allowed to improve and extended the previous correlation. A new correlation, called CHF-KFK-3, which accounts for these improvements, is presented in the paper. (orig.) [de

Experimental study of critical heat flux enhancement with hypervapotron structure under natural circulation conditions

Energy Technology Data Exchange (ETDEWEB)

Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Chang, Huajian [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhao, Yufeng, E-mail: zhaoyufeng@snptc.com.cn [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhang, Ming; Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei [State Power Investment Corporation, Beijing (China)

2017-05-15

Highlights: • Natural circulation tests are performed to study the effect of hypervapotron on CHF. • Hypervapotron structure improves CHF under natural circulation conditions. • Visualization data illustrate vapor blanket behavior under subcooled flow conditions. - Abstract: The enhancement of critical heat flux with a hypervapotron structure under natural circulation conditions is investigated in this study. Subcooled flow boiling CHF experiments are performed using smooth and hypervapotron surfaces at different inclination angles under natural circulation conditions. The experimental facility, TESEC (Test of External Vessel Surface with Enhanced Cooling), is designed to conduct CHF experiments in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. The two-phase flow of subcooled flow boiling on both smooth and hypervapotron heating plates is observed and analyzed by the high-speed visualization technology. The results show that both smooth surface and hypervapotron surface CHF data exhibit a similar trend against inclination angles compared with the CHF results under forced flow condition on the same facility in earlier studies. However, the CHF enhancement of the hypervapotron structure is evidently more significant than the one under forced flow conditions. The experiments also indicate that the natural flow rates are higher with hypervapotron structure. The initiation of CHF is analyzed under transient subcooling and flow rate conditions for both smooth and hypervapotron heating surfaces. An explanation is given for the significant enhancement effect caused by the hypervapotron surface under natural circulation conditions. The visualization data are exhibited to demonstrate the behavior of the vapor blanket at various inclination angles and on different surfaces. The geometric data of the vapor blanket are quantified by an image post-processing method. It is found that the thickness of the vapor blanket

Critical heat flux experiments for high conversion light water reactor, (3)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

1990-03-01

As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

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

International Nuclear Information System (INIS)

Feldman, E.E.

2008-01-01

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. (author)

Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

International Nuclear Information System (INIS)

Ohtake, H.; Koizumi, Y.; Takano, K.I.

2001-01-01

The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

An analytic model of pool boiling critical heat flux on an immerged downward facing curved surface

International Nuclear Information System (INIS)

He, Hui; Pan, Liang-ming; Wu, Yao; Chen, De-qi

2015-01-01

Highlights: • Thin liquid film and supplement of liquid contribute to the CHF. • CHF increases from the bottom to the upper of the lowerhead. • Evaporation of thin liquid film is dominant nearby bottom region. • The subcooling has significant effects on the CHF. - Abstract: In this paper, an analytical model of the critical heat flux (CHF) on the downward facing curved surface for pool boiling has been proposed, which hypothesizes that the CHF on the downward facing curved is composed of two parts, i.e. the evaporation of the thin liquid film underneath the elongated bubble adhering to the lower head outer surface and the depletion of supplement of liquid due to the relative motion of vapor bubbles along with the downward facing curved. The former adopts the Kelvin–Helmholtz instability analysis of vapor–liquid interface of the vapor jets which penetrating in the thin liquid film. When the heat flux closing to the CHF point, the vapor–liquid interface becomes highly distorted, which block liquid to feed the thin liquid film and the thin liquid film will dry out gradually. While the latter considers that the vapor bubbles move along with the downward facing curved surface, and the liquid in two-phase boundary layer enter the liquid film that will be exhausted when the CHF occurs. Based on the aforementioned mechanism and the energy balance between the thin liquid film evaporation and water feeding, and taking the subcooling of the bulk water into account, the mathematic model about the downward facing curved surface CHF has been proposed. The CHF of the downward facing curved surface for pool boiling increases along with the downward facing orientation except in the vicinity of bottom center region, because in this region the vapor bubble almost stagnates and the evaporation of the thin liquid film is dominant. In addition, the subcooling has significant effect on the CHF. Comparing the result of this model with the published experimental results show

Influence of test tube material on subcooled flow boiling critical heat flux in short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Shiotsu, Masahiro; Noda, Nobuaki

2007-01-01

The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u=4.0 to 13.3 m/s), the inlet subcoolings (ΔT sub,in =48.6 to 154.7 K), the inlet pressure (P in =735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tube of inner diameter (d=6 mm), heated length (L=66 mm) and L/d=11 with the inner surface of rough finished (Surface roughness, Ra=3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tube of d=6 mm, L=60 mm and L/d=10 with Ra=0.18 μm and the Platinum (Pt) test tubes of d=3 and 6 mm, L=66.5 and 69.6 mm, and L/d=22.2 and 11.6 respectively with Ra=0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (author)

Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube

International Nuclear Information System (INIS)

Koichi Hata; Masahiro Shiotsu; Nobuaki Noda

2006-01-01

The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling (ΔT sub,in = 48.6 to 154.7 K), the inlet pressure (P in = 735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R a = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R a = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R a = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)

Pool boiling CHF enhancement by micro/nanoscale modification of zircaloy-4 surface

International Nuclear Information System (INIS)

Ahn, Ho Seon; Lee, Chan; Kim, Hyungdae; Jo, HangJin; Kang, SoonHo; Kim, Joonwon; Shin, Jeongseob; Kim, Moo Hwan

2010-01-01

Consideration of the critical heat flux (CHF) requires difficult compromises between economy and safety in many types of thermal systems, including nuclear power plants. Much research has been directed towards enhancing the CHF, and many recent studies have revealed that the significant CHF enhancement in nanofluids is due to surface deposition of nanoparticles. The surface deposition of nanoparticles influenced various surface characteristics. This fact indicated that the surface wettability is a key parameter for CHF enhancement and so is the surface morphology. In this study, surface wettability of zircaloy-4 used as cladding material of fuel rods in nuclear power plants was modified using surface treatment technique (i.e. anodization). Pool boiling experiments of distilled water on the prepared surfaces was conducted at atmospheric and saturated conditions to examine effects of the surface modification on CHF. The experimental results showed that CHF of zircaloy-4 can be significantly enhanced by the improvement in surface wettability using the surface modification, but only the wettability effect cannot explain the CHF increase on the treated zircaloy-4 surfaces completely. It was found that below a critical value of contact angle (10 o ), micro/nanostructures created by the surface treatment increased spreadability of liquid on the surface, which could lead to further increase in CHF even beyond the prediction caused only by the wettability improvement. These micro/nanostructures with multiscale on heated surface induced more significant CHF enhancement than it based on the wettability effect, due to liquid spreadability.

Critical heat flux for free convection boiling in thin rectangular channels

International Nuclear Information System (INIS)

Cheng, Lap Y.; Tichler, P.R.

1991-01-01

A review of the experimental data on free convection boiling critical heat flux (CHF) in vertical rectangular channels reveals three mechanisms of burnout. They are the pool boiling limit, the circulation limit, and the flooding limit associated with a transition in flow regime from churn to annular flow. The dominance of a particular mechanism depends on the dimensions of the channel. Analytical models were developed for each free convection boiling limit. Limited agreement with data is observed. A CHF correlation, which is valid for a wide range of gap sizes, was constructed from the CHFs calculated according to the three mechanisms of burnout. 17 refs., 7 figs

Mechanistic modeling of CHF in forced-convection subcooled boiling

International Nuclear Information System (INIS)

Podowski, M.Z.; Alajbegovic, A.; Kurul, N.; Drew, D.A.; Lahey, R.T. Jr.

1997-05-01

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

Evaluation of subcooled critical heat flux correlations for tubes with and without internal twisted tapes

International Nuclear Information System (INIS)

Inasaka, F.; Nariai, H.

1996-01-01

Eleven correlations and models for critical heat flux (CHF) of subcooled flow boiling in water were evaluated. Both a direct substitution method (DSM) and a heat balance condition method (HBM) were compared in the evaluations. The HBM was recommended as a better prediction method in the present study. For straight tubes under uniform heating conditions, the correlations of the Gunther, Knoebel, modified Tong, W-2, and Tong-75, and also the Celata and Weisman-Pei models were confirmed to give reasonably good predictions. Among them, the Celata model was the best with respect to accuracy. For swirl flow under uniform heating conditions, Tong-75-I (involving modification of the water velocity parameter) and Nariai-Inasaka correlations were confirmed to give reasonably good predictions, even though their predictions were too low for the CHF under non-uniform heating conditions. (orig.)

Assessment of CHF characteristics at subcooled conditions for the CANDU CANFLEX bundle

International Nuclear Information System (INIS)

Onder, E.N.; Leung, L.K.H.

2011-01-01

An analysis has been performed to assess the Critical Heat Flux (CHF) characteristics for the CANFLEX bundle at subcooled conditions. CHF characteristics for CANDU bundles have been established from experiments using full-scale bundle simulators. These experiments covered flow conditions of interest to normal operation and postulated loss-of-flow and small break loss-of-coolant accidents. Experimental CHF values obtained from these experiments were applied to develop correlations for analyses of regional overpower protection and safety trips. These correlations are applicable to the saturated region in the reference uncrept channel and the slightly subcooled region in postulated high-creep channels. Expanding the CHF data to subcooled conditions facilitates the evaluation of the margin to dryout at upstream bundle locations, even though dryout occurrences are not anticipated there. In view of the lack of experimental data, the ASSERT-PV subchannel code has been applied to establish CHF values at low qualities and high subcoolings (thermodynamic qualities corresponding to -25%). These CHF values have been applied to extend the CHF correlation to the highly subcooled conditions. (author)

The steady-state modeling and optimization of a refrigeration system for high heat flux removal

International Nuclear Information System (INIS)

Zhou Rongliang; Zhang Tiejun; Catano, Juan; Wen, John T.; Michna, Gregory J.; Peles, Yoav; Jensen, Michael K.

2010-01-01

Steady-state modeling and optimization of a refrigeration system for high heat flux removal, such as electronics cooling, is studied. The refrigeration cycle proposed consists of multiple evaporators, liquid accumulator, compressor, condenser and expansion valves. To obtain more efficient heat transfer and higher critical heat flux (CHF), the evaporators operate with two-phase flow only. This unique operating condition necessitates the inclusion of a liquid accumulator with integrated heater for the safe operation of the compressor. Due to the projected incorporation of microchannels into the system to enhance the heat transfer in heat sinks, the momentum balance equation, rarely seen in previous vapor compression cycle heat exchangers modeling efforts, is utilized in addition to the mass and energy balance equations to capture the expected significant microchannel pressure drop witnessed in previous experimental investigations. Using the steady-state model developed, a parametric study is performed to study the effect of various external inputs on the system performance. The Pareto optimization is applied to find the optimal system operating conditions for given heat loads such that the system coefficient of performance (COP) is optimized while satisfying the CHF and other system operation constraints. Initial validation efforts show the good agreement between the experimental data and model predictions.

Prediction of Critical Heat Flux under Rolling Motion

Energy Technology Data Exchange (ETDEWEB)

Hwang, Jinseok; Lee, Yeongun; Park, Gooncherl [Seoul National Univ., Seoul (Korea, Republic of)

2013-05-15

The aim to this paper may be summarized as follows: identify the flow regime compare with existing void-quality relationship and void fraction at OAF derived from the vapor superficial velocity obtained by the churn-to annular flow criterion, develop and evaluate the correlation for accurate prediction of CHF ratio under rolling motion. Experimentally measured CHF results from the previous study were not well-predicted by existing CHF correlations developed for wide range of pressure under rolling motion in vertical tube. Specifically, existing correlations do not account for the dynamic motion parameter, such as tangential and centrifugal force. This study reviewed some existing correlation and experimental studies related to reduction and enhancement of CHF and heat transfer and flow behavior under heaving and rolling motion, and developed a CHF ratio correlation for upward flow vertical tube under rolling motion. Based upon dimensionless groups, equations and interpolation factor, an empirical CHF correlation has been developed which is consistent with experimental data for uniformly heated tubes internally cooled by R-134 under rolling motion. Flow regime was determined through the prediction method for annular flow. Non-dimensional number and function were decided by CHF mechanism of each region. Interaction of LFD and DNB regions is taken into account by means of power interpolation which is reflected void fraction at OAF. The suggested correlation predicted the CHF Ratio with reasonable accuracy, showing an average error of -0.59 and 2.51% for RMS. Rolling motion can affect bubble motion and liquid film behavior complexly by combination of tangential and centrifugal forces and mass flow than heaving motion. Through a search of literature and a comparison of previous CHF ratio results, this work can contribute to the study of boiling heat transfer and CHF for the purpose of enhancement or reduction the CHF of dynamic motion system, such as marine reactor.

A critical heat flux correlation for advanced pressurized light water reactor application

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-05-01

Many CHF-correlations have been developed for water cooled rod clusters representing typical PWR or BWR fuel element geometries with relative wide rod lattices. However the fuel elements of an Advanced Pressurized Water Reactor (APWR) have a tight fuel rod lattice, in view of increasing the fuel utilization. It was therefore decided to produce a new CHF-correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry dependent parameters of this correlation were determined again with the method of the root mean square fitting from the experimental data of the CHF-tests performed in the frame of the Light Water Breeder Reactor programme at the Bettis Laboratory. These tests include triangular array rod bundles with very tight lattices. Furthermore the effect of spiral spacer ribs was investigated on the basis of experimental data from the Columbia University. Application of the new CHF-correlation to conditions typical for an APWR shows that the predicted critical heat fluxes are much smaller than those calculated with the usual PWR-CHF-correlations, but they are higher than those predicted by the B+W-VPI+SU correlation. (orig.) [de

Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

International Nuclear Information System (INIS)

Kiteley, J.C.; Carver, M.B.; Zhou, Q.N.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

Current Needs for the Experimental Investigation of the CHF Phenomenon Relevant to LWR Core Conditions

International Nuclear Information System (INIS)

Le Corre, J.M.

2009-01-01

The current achievements and needs toward the investigation, understanding and mechanistic prediction of the Critical Heat Flux (CHF) event, under PWR and BWR core conditions, are addressed in this paper. It is shown that, even when using advanced 3-D CFD simulation tools, the current approach to CHF mechanistic modeling has serious limitations. This is mainly due to the lack of information regarding the relevant two-phase flow pattern(s) (in particular near the heated wall) and associated mechanisms (at the meso and micro-scale) leading to the CHF event. Areas of current experimental needs are identified in order to address these shortcomings. In addition, the use of 1-D and 3-D numerical tools to mechanistically predict the CHF is discussed. It is shown that 3-D two-phase CFD codes may not be superior to 1-D codes without proper consideration of relevant constitutive relations. (author)

Mechanism of subcooled water flow boiling critical heat flux in a circular tube at high liquid Reynolds number

International Nuclear Information System (INIS)

Hata, K.; Fukuda, K.; Masuzaki, S.

2014-01-01

The subcooled boiling heat transfer and the steady state critical heat flux (CHF) in a vertical circular tube for the flow velocities (u=3.95 to 30.80 m/s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d=6 mm) and heated length (L=59.5 mm) is used in this work. The outer surface temperatures of the SUS304 test tube with heating are observed by an infrared thermal imaging camera and a video camera. The subcooled boiling heat transfers for SUS304 test tube are compared with the values calculated by other workers' correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details based on the experimental data. Nucleate boiling surface superheats at the CHF are close to the lower limit of the heterogeneous spontaneous nucleation temperature and the homogeneous spontaneous nucleation temperature. The dominant mechanism of the subcooled flow boiling CHF on the SUS304 circular tube is discussed at high liquid Reynolds number. On the other hand, theoretical equations for k-ε turbulence model in a circular tube of a 3 mm in diameter and a 526 mm long are numerically solved for heating of water on heated section of a 3 mm in diameter and a 67 mm long with various thicknesses of conductive sub-layer by using PHOENICS code under the same conditions as the experimental ones previously obtained considering the temperature dependence of thermo-physical properties concerned. The Platinum (Pt) test tube of inner diameter (d=3 mm) and heated length (L=66.5 mm) was used in this experiment. The thicknesses of conductive sub-layer from non-boiling regime to CHF are clarified. The thicknesses of conductive sub-layer at the CHF point are evaluated for various flow velocities. The experimental values of the CHF are also compared with the corresponding

Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

International Nuclear Information System (INIS)

Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

2001-09-01

Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

Development of an artificial neural network to predict critical heat flux based on the look up tables

Energy Technology Data Exchange (ETDEWEB)

Terng, Nilton; Carajilescov, Pedro, E-mail: Nil.terng@gmail.com, E-mail: pedro.carajilescov@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais

2015-07-01

The critical heat flux (CHF) is one of the principal thermal hydraulic limits of PWR type nuclear reactors. The present work consists in the development of an artificial neural network (ANN) to estimate the CHF, based on Look Up Table CHF data, published by Groeneveld (2006). Three parameters were considered in the development of the ANN: the pressure in the range of 1 to 21 MPa, the mass flux in the range of 50 to 8000 kg m{sup -2} s{sup -1} and the thermodynamic quality in the range of - 0.5 to 0.9. The ANN model considered was a multi feed forward net, which have two feedforward ANN. The first one, called main neural network, is used to calculate the result of CHF, and the second, denominated spacenet, is responsible to modify the main neural network according to the input. Comparing the ANN predictions with the data of the Look Up Table, it was observed an average of the ratio of 0.993 and a root mean square error of 13.3%. With the developed ANN, a parametric study of CHF was performed to observe the influence of each parameter in the CHF. It was possible to note that the CHF decreases with the increase of pressure and thermodynamic quality, while CHF increases with the mass flow rate, as expected. However, some erratic trends were also observed which can be attributed to either unknown aspect of the CHF phenomenon or uncertainties in the data. (author)

Effect of the Aligned Flow Obstacles on Downward-Facing CHF in an Inclined Rectangular Channel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Ui ju; Son, Hong Hyun; Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong [Hanyang, Seoul (Korea, Republic of)

2016-05-15

The cooling channel consists of the inclined (10 .deg. ) portion of the downward facing heating channel and vertical portion of the heating channel. Features unique to flow boiling with the downward-facing heater surface in the inclined cooling channel where the studs are installed have drawn a considerable attention. That's because prior studies on boiling crisis indicate the orientation of the heated wall can exert substantial influence on CHF. Especially, the concentration of the vapor near the downward facing heater surface makes this region susceptible to premature boiling crisis when compared to vertical or upward-facing heated wall. Also, the installed studs could cause a partial flow blockage, and distort the flow streamline. Due to the distortion, stagnation points may occur in the cooling channel, promoting the concentration of the vapor near the heated wall. Then, the locally degraded heat transfer around the points may result in the formation of vapor pocket. The primary objective of this study is to make available experimental data on the CHF values varying the shape of studs and to improve understanding of the mechanism of flow boiling crisis associated with the aligned flow obstructions by means of visual experimental study. This study presents experimental data for subcooled flow boiling of water at atmospheric pressure and low mass flux conditions. The major outcomes from this investigation can be summarized as follows: (1) The CHF value from bare test section is -320kW/m{sup 2} , significantly lower than the values from the existing correlations even considering the uncertainty in the experiments. (2) The CHF value is remarkably decreased as columnar structures are installed in the channel. It is confirmed that formation and extinction of local dryout occurs repeatedly just behind the first stud at heat flux of -160 kW/m{sup 2}.

Critical heat flux and flow instability in an advanced light water reactor

International Nuclear Information System (INIS)

Dae-Hyun Hwang; Kyong-Won Seo; Chung-Chan Lee; Sung-Kyun Zee

2005-01-01

Full text of publication follows: An advanced light water reactor concept has been continuously studied in KAERI with an output in the range of about 60 to 300 MW th . The reactor is purposed to be utilized as an energy source for seawater desalination as well as small scale power generation. In order to achieve the intrinsic safety and enhanced operational flexibility, some specific design considerations such as low power density and soluble boron free operation have been incorporated in the multiple-parallel-channel type reactor core. The low power density can be achieved by adopting fuel assemblies with tightly spaced non-square lattice rod array. The allowable core operating region should be primarily limited by the two design parameters; the critical heat flux(CHF) and the flow instabilities in the multiple parallel fuel assembly channels. The characteristics of CHF and flow instability have been investigated through experimental and analytical works. The CHF prediction model was established on the basis of experimental data obtained from 19-rod test bundles. The CHF experiments have been conducted for various test bundles with different heated lengths, uniform and non-uniform radial and axial power distributions, water and Freon as the working fluids, and different number of unheated rods. The parametric ranges of CHF experiments covers the pressure from 6 to 18 MPa, the mass flux from 150 to 2000 kg/m 2 /s, and the inlet subcooling from 10 to 120 deg. C. The flow instabilities due to density wave oscillations were investigated by conducting experiments with two parallel channels under the pressure ranges from 6 to 16 MPa. The parametric behavior of flow instability was examined for the test sections with different lengths of adiabatic risers, different axial power shapes, different inlet restrictions, and different channel cross sections. The stability boundary was experimentally determined by increasing channel inlet temperature or reducing the flow rate

Measurement of critical heat flux in narrow gap with two-dimensional slices

International Nuclear Information System (INIS)

Kim, Yong Hoon; Kim, Sung Joong; Noh, Sang Woo; Suh, Kune Y.

2002-01-01

A cooling mechanism due to boiling in a gap between the debris crust and the reactor pressure vessel (RPV) wall was proposed for the TMI-2 reactor accident analysis. If there is enough heat transfer through the gap to cool the outer surface of the debris and the inner surface of the wall, the RPV wall may preserve its integrity during a severe core melt accident. If the heat removal through gap cooling relative to the counter-current flow limitation (CCFL) is pronounced, the safety margin of the reactor can be far greater than what had been previously known in the severe accident management arena. Should a severe accident take place, the RPV integrity will be maintained because of the inherent nature of degraded core coolability inside the lower head due to boiling in a narrow gap between the debris crust and the RPV wall. As a defense-in-depth measure, the heat removal capability by gap cooling coupled with external cooling can be examined for the Korean Standard Nuclear Power Plant (KSNPP) and the Advanced Power Reactor 1400MWe (APR1400) in light of the TMI-2 vessel survival. A number of studies were carried out to investigate the complex heat transfer mechanisms for the debris cooling in the lower plenum. However, these heat transfer mechanisms have not been clearly understood yet. The CHFG (Critical Heat Flux in Gap) experiments at KAERI were carried out to develop the critical heat flux (CHF) correlation in a hemispherical gap, which is the upper limit of the heat transfer. According to the CHFG experiments performed with a pool boiling condition, the CHF in a parallel gap was reduced by 1/30 compared with the value measured in the open pool boiling condition. The correlation developed from the CHFG experiment is based on the fact that the CHF in a hemispherical gap is governed by the CCFL and a Kutateladze type CCFL parameter correlates CCFL data well in hemispherical gap geometry. However, the results of the CHFG experiments appear to be limited in their

Particle swarm optimization-based least squares support vector regression for critical heat flux prediction

International Nuclear Information System (INIS)

Jiang, B.T.; Zhao, F.Y.

2013-01-01

Highlights: ► CHF data are collected from the published literature. ► Less training data are used to train the LSSVR model. ► PSO is adopted to optimize the key parameters to improve the model precision. ► The reliability of LSSVR is proved through parametric trends analysis. - Abstract: In view of practical importance of critical heat flux (CHF) for design and safety of nuclear reactors, accurate prediction of CHF is of utmost significance. This paper presents a novel approach using least squares support vector regression (LSSVR) and particle swarm optimization (PSO) to predict CHF. Two available published datasets are used to train and test the proposed algorithm, in which PSO is employed to search for the best parameters involved in LSSVR model. The CHF values obtained by the LSSVR model are compared with the corresponding experimental values and those of a previous method, adaptive neuro fuzzy inference system (ANFIS). This comparison is also carried out in the investigation of parametric trends of CHF. It is found that the proposed method can achieve the desired performance and yields a more satisfactory fit with experimental results than ANFIS. Therefore, LSSVR method is likely to be suitable for other parameters processing such as CHF

A theoretical prediction of critical heat flux in subcooled pool boiling during power transients

International Nuclear Information System (INIS)

Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

1988-01-01

Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions are of fundamenatal interest in the design, operation, safety of boiling and two-phase flow devices. This paper discusses the results of a comprehensive theoretical study made specifically to model transient CHF behavior in subcooled pool boiling. This study is based upon a simplified steady-state CHF model in terms of the vapor mass growth period. The results obtained from this theory indicate favorable agreement with the experimental data from cylindrical heaters with small radii. The statistical nature of the vapor mass behavior in transient boiling also is considered and upper and lower limits for the current theory are established. Various factors that affect the discrepancy between the data and the theory are discussed

Theoretical modeling of CHF for near-saturated pool boiling and flow boiling from short heaters using the interfacial lift-off criterion

International Nuclear Information System (INIS)

Mudawar, I.; Galloway, J.E.; Gersey, C.O.

1995-01-01

Pool boiling and flow boiling were examined for near-saturated bulk conditions in order to determine the critical heat flux (CHF) trigger mechanism for each. Photographic studies of the wall region revealed features common to both situations. At fluxes below CHF, the vapor coalesces into a wavy layer which permits wetting only in wetting fronts, the portions of the liquid-vapor interface which contact the wall as a result of the interfacial waviness. Close examination of the interfacial features revealed the waves are generated from the lower edge of the heater in pool boiling and the heater's upstream region in flow boiling. Wavelengths follow predictions based upon the Kelvin-Helmholtz instability criterion. Critical heat flux in both cases occurs when the pressure force exerted upon the interface due to interfacial curvature, which tends to preserve interfacial contact with the wall prior to CHF, is overcome by the momentum of vapor at the site of the first wetting front, causing the interface to lift away from the wall. It is shown this interfacial lift-off criterion facilitates accurate theoretical modeling of CHF in pool boiling and in flow boiling in both straight and curved channels

Theoretical modeling of CHF for near-saturated pool boiling and flow boiling from short heaters using the interfacial lift-off criterion

Energy Technology Data Exchange (ETDEWEB)

Mudawar, I.; Galloway, J.E.; Gersey, C.O. [Purdue Univ., West Lafayette, IN (United States)] [and others

1995-12-31

Pool boiling and flow boiling were examined for near-saturated bulk conditions in order to determine the critical heat flux (CHF) trigger mechanism for each. Photographic studies of the wall region revealed features common to both situations. At fluxes below CHF, the vapor coalesces into a wavy layer which permits wetting only in wetting fronts, the portions of the liquid-vapor interface which contact the wall as a result of the interfacial waviness. Close examination of the interfacial features revealed the waves are generated from the lower edge of the heater in pool boiling and the heater`s upstream region in flow boiling. Wavelengths follow predictions based upon the Kelvin-Helmholtz instability criterion. Critical heat flux in both cases occurs when the pressure force exerted upon the interface due to interfacial curvature, which tends to preserve interfacial contact with the wall prior to CHF, is overcome by the momentum of vapor at the site of the first wetting front, causing the interface to lift away from the wall. It is shown this interfacial lift-off criterion facilitates accurate theoretical modeling of CHF in pool boiling and in flow boiling in both straight and curved channels.

Accuracy improvement of SPACE code using the optimization for CHF subroutine

International Nuclear Information System (INIS)

Yang, Chang Keun; Kim, Yo Han; Park, Jong Eun; Ha, Sang Jun

2010-01-01

Typically, a subroutine to calculate the CHF (Critical Heat Flux) is loaded in code for safety analysis of nuclear power plant. CHF subroutine calculates CHF phenomenon using arbitrary condition (Temperature, pressure, flow rate, power, etc). When safety analysis for nuclear power plant is performed using major factor, CHF parameter is one of the most important factor. But the subroutines used in most codes, such as Biasi method, etc., estimate some different values from experimental data. Most CHF subroutines in the codes could predict only in their specification area, such as pressure, mass flow, void fraction, etc. Even though the most accurate CHF subroutine is used in the high quality nuclear safety analysis code, it is not assured that the valued predicted values by the subroutine are acceptable out of their application area. To overcome this hardship, various approaches to estimate the CHF have been examined during the code developing stage of SPACE. And the six sigma technique was adopted for the examination as mentioned this study. The objective of this study is to improvement of CHF prediction accuracy for nuclear power plant safety analysis code using the CHF database and Six Sigma technique. Through the study, it was concluded that the six sigma technique was useful to quantify the deviation of prediction values to experimental data and the implemented CHF prediction method in SPACE code had well-predict capabilities compared with those from other methods

Critical heat flux prediction by using radial basis function and multilayer perceptron neural networks: A comparison study

International Nuclear Information System (INIS)

Vaziri, Nima; Hojabri, Alireza; Erfani, Ali; Monsefi, Mehrdad; Nilforooshan, Behnam

2007-01-01

Critical heat flux (CHF) is an important parameter for the design of nuclear reactors. Although many experimental and theoretical researches have been performed, there is not a single correlation to predict CHF because it is influenced by many parameters. These parameters are based on fixed inlet, local and fixed outlet conditions. Artificial neural networks (ANNs) have been applied to a wide variety of different areas such as prediction, approximation, modeling and classification. In this study, two types of neural networks, radial basis function (RBF) and multilayer perceptron (MLP), are trained with the experimental CHF data and their performances are compared. RBF predicts CHF with root mean square (RMS) errors of 0.24%, 7.9%, 0.16% and MLP predicts CHF with RMS errors of 1.29%, 8.31% and 2.71%, in fixed inlet conditions, local conditions and fixed outlet conditions, respectively. The results show that neural networks with RBF structure have superior performance in CHF data prediction over MLP neural networks. The parametric trends of CHF obtained by the trained ANNs are also evaluated and results reported

Development of heat transfer models for gap cooling

Energy Technology Data Exchange (ETDEWEB)

Kohriyama, Tamio; Murase, Michio; Tamaki, Tomohiko [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

In a severe accident of a light water reactor (LWR), heat transfer models in a narrow annular gap between superheated core debris and a reactor pressure vessel (RPV) are important to evaluate the integrity of RPV and emergency procedures. This paper discusses the effects of superheat on the heat flux based on existing data. In low superheat conditions, the heat flux in the narrow gap is higher than the heat flux in pool nucleate boiling due to restricted flow area. It approaches the nucleate boiling heat flux as superheat increasing and reaches a critical value subject to the counter-current flow limiting (CCFL) at the top end of the gap. A heat transfer correlation was derived as a function of dimensionless superheat and a Kutateladze-type CCFL correlation was deduced for critical heat flux (CHF) restricted by CCFL, which gave good prediction for a wide range of the CHF data. Effect of an angle of inclination of the gap could also be incorporated in the CCFL correlation. In high superheat conditions, the heat flux in the narrow gap maintains a similar shape to the pool boiling curve but shifts the position to a higher superheated side than the pool boiling except film boiling, which could be expressed by the typical pool film boiling correlation. Incorporating quench test data, the heat flux correlation was derived as a function of dimensionless superheat using the same formula for the low superheat and the Kutateladze-type CCFL correlation was deduced for CHF. The CHF at the high superheat was 3-4 times as large as CHF at the low superheat and this difference was well predicted by different flow patterns in the gap and the balance of pressure gradients between gas and liquid phases. (author)

Analysis of in-R12 CHF data: influence of hydraulic diameter and heating length; test of Weisman boiling crisis model; Analyse de donnees de flux critique en R12: influence du diametre hydraulique et de la longueur chauffante; test du modele de Weisman

Energy Technology Data Exchange (ETDEWEB)

Czop, V; Herer, C; Souyri, A; Garnier, J

1993-09-01

In order to progress on the comprehensive modelling of the boiling crisis phenomenon, Electricite de France (EDF), Commissariat a l`Energie Atomique (CEA) and FRAMATOME have set up experimental programs involving in-R12 tests: the EDF APHRODITE program and the CEA-EDF-FRAMATOME DEBORA program. The first phase in these programs aims to acquire critical heat flux (CHF) data banks, within large thermal-hydraulic parameter ranges, both in cylindrical and annular configurations, and with different hydraulic diameters and heating lengths. Actually, three data banks have been considered in the analysis, all of them concerning in-R12 round tube tests: - the APHRODITE data bank, obtained at EDF with a 13 mn inside diameter, - the DEBORA data bank, obtained at CEA with a 19.2 mm inside diameter, - the KRISTA data bank, obtained at KfK with a 8 mm inside diameter. The analysis was conducted using CHF correlations and with the help of an advanced mathematical tool using pseudo-cubic thin plate type Spline functions. Two conclusions were drawn: -no influence of the heating length on our CHF results, - the influence of the diameter on the CHF cannot be simply expressed by an exponential function of this parameter, as thermal-hydraulic parameters also have an influence. Some calculations with Weisman and Pei theoretical boiling crisis model have been compared to experimental values: fairly good agreement was obtained, but further study must focus on improving the modelling of the influence of pressure and mass velocity. (authors). 12 figs., 4 tabs., 21 refs.

Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

Energy Technology Data Exchange (ETDEWEB)

Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

2013-12-15

Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

International Nuclear Information System (INIS)

Park, Seong Dae; Bang, In Cheol

2013-01-01

Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m 2 s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface

Theoretical and experimental investigations of CHF in round tubes and rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun

1994-02-01

A knowledge of the condition leading to critical heat flux (CHF) is of great importance in the design of nuclear reactors. Although many efforts have been devoted to the subject of CHF during the last few decades, information on the burnout phenomenon at low velocity condition is very limited. Furthermore, in most cases, the applicable range of a bundle CHF correlation is restricted to a narrow region mainly due to the limitation of the CHF data base used in the correlation development. In view of these points, theoretical and experimental investigations are performed in this study for round tubes and rod bundles. A CHF prediction model for low velocity conditions is proposed throughout the assessment of CHF data from various sources with mass velocities less than 500 kg/m 2 s. The CHF data base is classified into seven groups with respect to the flow pattern characteristics at CHF conditions. CHF data for each group is analyzed by several CHF prediction models including; the flooding correlations, the flow regime transition criteria, the complete evaporation model, and the empirical correlations. At zero inlet flow or extremely low mass velocity conditions, the flooding correlation can be used for predicting CHF employing appropriate constant. In the slug or churn-turbulent flow regime, CHF seems to occur at the annular flow transition conditions. When CHF occurs at the annular flow region, the empirical correlation such as AECL CHF lookup table gives accurate predictions except for the ranges where density-wave instability is expected. A phenomenological model for the prediction of dryout locations under flooding-limited CHF condition is developed based on the liquid film dryout model and the two-phase mixture level theory. The mass and energy conservation equations are applicable to the liquid film considering no entrainment of liquid droplets from the film region. The variation of the two-phase mixture level after the onset of flooding is calculated based on

Investigation on premature occurrence of critical heat flux under oscillatory flow and power conditions

International Nuclear Information System (INIS)

Vishnoi, A.K.; Dasgupta, A.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

2015-01-01

Two-phase natural circulation loops have extensive applications in nuclear and process industries. One of the major concerns with natural circulation is the occurrence of the various types of flow instabilities, which can cause premature boiling crisis due to flow and power oscillations. In this work a transient computer code COPCOS (Code for Prediction of CHF under Oscillating flow and power condition) has been developed to predict the premature occurrence of CHF (critical heat flux) under oscillating flow and power. The code incorporates conduction equation of the fuel and coolant energy equation. For CHF prediction, CHF look-up table developed by Groeneveld is used. A facility named CHF and Instability Loop (CHIL) has been set up to study the effect of oscillatory flow on CHF. CHF and Instability Loop (CHIL) is a simple rectangular loop having a 10.5 mm ID and 1.2 m long test section. The flow through the test section is controlled by a canned motor pump using a Variable Frequency Drive (VFD). This leads to the ability of having a very precise control over flow oscillations which can be induced in the test section. The effect of frequency and amplitude of flow oscillation on occurrence of premature CHF has been investigated in this facility using COPCOS. Full paper covers details of COPCOS code, description of the facility and effect of frequency and the effect of oscillatory flow on CHF in the facility. (author)

General correlation for prediction of critical heat flux ratio in water cooled channels

Energy Technology Data Exchange (ETDEWEB)

Pernica, R.; Cizek, J.

1995-09-01

The paper present the general empirical Critical Heat Flux Ration (CHFR) correlation which is valid for vertical water upflow through tubes, internally heated concentric annuli and rod bundles geometries with both wide and very tight square and triangular rods lattices. The proposed general PG correlation directly predicts the CHFR, it comprises axial and radial non-uniform heating, and is valid in a wider range of thermal hydraulic conditions than previously published critical heat flux correlations. The PG correlation has been developed using the critical heat flux Czech data bank which includes more than 9500 experimental data on tubes, 7600 data on rod bundles and 713 data on internally heated concentric annuli. Accuracy of the CHFR prediction, statistically assessed by the constant dryout conditions approach, is characterized by the mean value nearing 1.00 and the standard deviation less than 0.06. Moverover, a subchannel form of the PG correlations is statistically verified on Westinghouse and Combustion Engineering rod bundle data bases, i.e. more than 7000 experimental CHF points of Columbia University data bank were used.

An experimental study on critical heat flux in a hemispherical narrow gap

International Nuclear Information System (INIS)

Park, R.J.; Lee, S.J.; Kang, K.H.; Kim, J.H.; Kim, S.B.; Kim, H.D.; Jeong, J.H.

2000-01-01

An experimental study of CHFG (Critical Heat Flux in Gap) has been performed to investigate the inherent cooling mechanism using distilled water and Freon R-113 in hemispherical narrow gaps. As a separate effect test of the CHFG test, a CCFL (Counter Current Flow Limit) test has been also performed to confirm the mechanism of the CHF in narrow annular gaps with large diameter. The CHFG test results have shown that an increase in the gap thickness leads to an increase in critical power. The pressure effect on the critical power was found to be much milder than predictions by CHF correlations of other studies. In the CCFL experiment, the occurrence of CCFL was correlated with the Wallis parameter, which was assumed to correspond to the critical power in the CHFG experiment. The measured values of critical power in the CHFG tests are much lower than CCFL experimental data and the predictions made by empirical CHF correlations. (author)

IR-thermography-based investigation of critical heat flux in subcooled flow boiling of water at atmospheric and high pressure conditions

Energy Technology Data Exchange (ETDEWEB)

Bucci, Matteo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Seong, Jee H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jdacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Richenderfer, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kossolapov, A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

2017-11-01

Here we report on MIT’s THM work in Q4 2016 and Q1 2017. The goal of this project is to design, construct and execute tests of flow boiling critical heat flux (CHF) at high-pressure using high-resolution and high-speed video and infrared (IR) thermometry, to generate unique data to inform the development of and validate mechanistic boiling heat transfer and CHF models. In FY2016, a new test section was designed and fabricated. Data was collected at atmospheric conditions at 10, 25 and 50 K subcoolings, and three mass fluxes, i.e. 500, 750 and 1000 kg/m2/s. Starting in Q4 2016 and continuing forward, new post-processing techniques have been developed to analyze the data collected. These new algorithms analyze the time-dependent temperature and heat flux distributions to calculate nucleation site density, nucleation frequency, growth and wait time, dry area fraction, and the complete heat flux partitioning. In Q1 2017 a new flow boiling loop was designed and constructed to support flow boiling tests up 10 bar pressure and 180 °C. Initial shakedown and testing has been completed. The flow loop and test section are now ready to begin high-pressure flow boiling testing.

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

International Nuclear Information System (INIS)

Feldman, E.

2008-01-01

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

Evaluation of mechanistic DNB models using HCLWR CHF data

International Nuclear Information System (INIS)

Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio.

1992-03-01

An onset of departure from nucleate boiling (DNB) in light water reactor (LWR) has been generally predicted with empirical correlations. Since these correlations have less physical bases and contain adjustable empirical constants determined by best fitting of test data, applicable geometries and flow conditions are limited within the original experiment ranges. In order to obtain more universal prediction method, several mechanistic DNB models based on physical approaches have been proposed in recent years. However, the predictive capabilities of mechanistic DNB models have not been verified successfully especially for advanced LWR design purposes. In this report, typical DNB mechanistic models are reviewed and compared with critical heat flux (CHF) data for high conversion light water reactor (HCLWR). The experiments were performed using triangular 7-rods array with non-uniform axial heat flux distribution. Test pressure was 16 MPa, mass velocities ranged from 800 t0 3100 kg/s·m 2 and exit qualities from -0.07 to 0.19. The evaluated models are: 1) Wisman-Pei, 2) Chang-Lee, 3) Lee-Mudawwar, 4) Lin-Lee-Pei, and 5) Katto. The first two models are based on near-wall bubble crowding model and the other three models on sublayer dryout model. The comparison with experimental data indicated that the Weisman-Pei model agreed relatively well with the CHF data. Effects of empirical constants in each model on CHF calculation were clarified by sensitivity studies. It was also found that the magnitudes of physical quantities obtained in the course of calculation were significantly different for each model. Therefore, microscopic observation of the onset of DNB on heated surface is essential to clarify the DNB mechanism and establish a general DNB mechanistic model based on physical phenomenon. (author)

Thermal and mechanical behavior of APWR-claddings under critical heat flux conditions

International Nuclear Information System (INIS)

Diegele, E.; Rust, K.

1986-10-01

Helical grid spacers, such as three or six helical fins as integral part of the claddings, are regarded as a more convenient design for the very tight lattice of an advanced pressurized water reactor (APWR) than grid spacers usually used. Furthermore, it is expected that this spacer design allows an increased safety margin against the critical heat flux (CHF), the knowledge of which is important for design, licensing, and operation of water cooled reactors. To address the distribution of the heat flux density at the outer circumference of the cladding geometry under investigation, the temperature fields in claddings without as well with fins were calculated taking into consideration nuclear and electrically heated rods. Besides the thermal behavior of the claddings, the magnitude and distribution of thermal stresses were determined additionally. A locally increased surface heat flux up to about 40 percent was calculated for the fin bases of nuclear as well as indirect electrically heated claddings with six such helical fins. For all investigated cases, the VON MISES stresses are clearly lower than 200 MPa, implying that no plastic deformations are to be expected. The aim of this theoretical analysis is to allow a qualitative assessment of the finned tube conception and to support experimental investigations concerning the critical heat flux. (orig.) [de

Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

International Nuclear Information System (INIS)

Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

2011-01-01

The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

Experimental study on the effect of gap size to CCFL and CHF in a vertical of narrow rectangular channel during quenching process

International Nuclear Information System (INIS)

Juarsa, Mulya; Putra, Nandy; Septiadi, Wayan Nata; Antariksawan, Anhar Riza

2014-01-01

Highlights: • Quenching in narrow rectangular channel with gap sizes variation was investigated. • The mechanism of counter-current flow depends on gap sizes variation. • The results confirmed the existence of CCFL in narrow rectangular channels. • CHF and mass flux gradient in the quenching was about 0.22 times than steady state. • Modification of CHF and mass flow rate dimensionless correlation was established. - Abstract: The quenching process has become an important thermal management study to intensify the safety margin for the integrity of the reactor vessel under the core meltdown condition. The boiling heat transfer mechanism in the channel is one aspect that needs further examination. The present study aimed to investigate the effect of the differences in channel gap size to counter-current flow limitation (CCFL) and critical heat flux (CHF) during transient cooling in atmospheric pressure and quenching using two vertical plates with 1 mm, 2 mm, and 3 mm gap sizes and heated length of 1100 mm. The initial temperature of the plate was set at 600 °C. Cooling water mass flow rate and sib-cooled temperature were set at about 0.089 kg/s and 90 °C, respectively. Calculations were performed to obtain the CHF value through the boiling curve using transient temperature data. Non-dimensional correlations from other research study was used in this research. The influence of gap sizes on CCFL and CHF resulted in an increased value of CHF relative to gap size; additionally, the CHF for gap sizes of 2 mm and 3 mm increased about 34.4% and 140.5%, respectively, compared to the CHF for the 1 mm gap size. In this research, a curve map of the relationship between non-dimensional CHF and non-dimensional mass flux of water flowing downward shows that the correlation of this experimental study has a gradient number of about 0.22 similar to Mishima and Nishihara correlation. The results confirmed the existence of CCFL in the vertical narrow rectangular channels due

Neutralized wettability effect of superhydrophilic Cr-layered surface on pool boiling critical heat flux

International Nuclear Information System (INIS)

Son, Hong Hyun; Jeong, Ui Ju; Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

2016-01-01

The former method is deemed challenging due to longer development period and license issue. In this regard, FeCrAl, Cr, and SiC have been received positive attention as ATF coating materials because they are highly resistant to high temperature steam reaction causing massive hydrogen generation. In this study, Cr was selected as a target deposition material on the metal substrate because we found that Cr-layered surface becomes superhydrophilic, favorable to delaying the triggering of the critical heat flux (CHF). Thus in order to investigate the effect of Cr-layered superhydrophilic surfaces (under explored coating conditions) on pool boiling heat transfer, pool boiling experiment was conducted in the saturated deionized water under atmospheric pressure. As a physical vapor deposition (PVD) method, the DC magnetron sputtering technique was introduced to develop Cr-layered nanostructure. As a control variable of DC sputtering, substrate temperature was selected. Surface wettability and nanostructure were analyzed as major surface parameters on the CHF. We believe that highly dense micro/nano structure without nucleation cavities and inner pores neutralized the wettability effect on the CHF. Moreover, superhydrophilic surface with deficient cavity density rather hinders active nucleation. This emphasizes the importance of micro/nano structure surface for enhanced boiling heat transfer.

Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

2003-05-01

Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

Heat transfer phenomena related to the boiling crisis

International Nuclear Information System (INIS)

Groenveld, D.C.

1981-03-01

This report contains a state-of-the-art review of critical heat flux (CHF) and post-CHF heat transfer. Part I reviews the mechanisms controlling the boiling crisis. The observed parametric trends of the CHF in a heat flux controlled system are discussed in detail, paying special attention to parameters pertaining to nuclear fuel. The various methods of predicting the critical power are described. Part II reviews the published information on transition boiling and film boiling heat transfer under forced convective conditions. Transition boiling data were found to be available only within limited ranges of conditions. The data did not permit the derivation of a correlation; however, the parametric trends were isolated from these data. (author)

Steady-state and transient studies on critical heat flux of a PWR 5 x 5 fuel element bundle with complex spacer wire geometry

International Nuclear Information System (INIS)

Fulfs, H.; Katsaounis, A.; Kreubig, M.; Minden, C. von; Orlowski, R.

1980-01-01

The results will be described in exemplary presentations completely and concluding. The experimental examination of the steady state simularity of critical heat flux (CHF) in freon 12 and water at identical PWR-5 x 15-rod bundles will show that hot rod/hot channels position as well as CHF can be transformed from model to original fluid with good accuracy. The investigated mass flow and power transients (only in freon 12) point out a definite influence of initial and boundary conditions on CHF and CHF time delay at changing rates higher than 10 to 20%/s. On the contrary simulation of primary pump failure (LOFA) shows no or only small improvement in CHF behaviour while a coupled Scram prevents from reaching the boiling crisis. (orig.) [de

Overview PWR-Blowdown Heat Transfer Separate-Effects Program

International Nuclear Information System (INIS)

White, J.D.

1978-01-01

The ORNL Pressurized Water Reactor Blowdown Heat Transfer Program (PWR-BDHT) is a separate-effects experimental study of thermal-hydraulic phenomena occurring during the first 20 sec of a hypothetical LOCA. Specific objectives include the determination, for a wide range of parameters, of time to CHF and the following variables for both pre- and post-CHF: heat fluxes, ΔT (temperature difference between pin surface and fluid), heat transfer coefficients, and local fluid properties. A summary of the most interesting results from the program obtained during the past year is presented. These results are in the area of: (1) RELAP verification, (2) electric pin calibration, (3) time to critical heat flux (CHF), (4) heat transfer coefficient comparisons, and (5) nuclear fuel pin simulation

Enhanced pool boiling critical heat flux induced by capillary wicking effect of a Cr-sputtered superhydrophilic surfaces

Energy Technology Data Exchange (ETDEWEB)

Son, Hong Hyun; Seo, Gwang Hyeok; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

In light of boiling heat transfer, the smooth surface potentially reduces active nucleation of bubbles and rewetting of dry spots near the critical heat flux (CHF). This kind of process is highly likely to deteriorate the CHF. Thus, it is essential to produce appropriate microstructures on the surface for the enhancement of the CHF. In this study, to investigate the microstructural effect of thin film-fabricated surfaces on the pool boiling CHF, we controlled the surface roughness in a narrow range of 0.1-0.25 μm and its morphologies, in the form of micro-scratches using PVD sputtering technique. Specifically for DC magnetron sputtering, pure chromium (Cr) was selected as a target material owing to its high oxidation resistance. In order to analyze the CHF trend with changes in roughness, we introduced existing capillary wicking-based models because superhydrophilic characteristics of microstructures are highly related to the capillary wicking behaviors in micro-flow channels. After Cr sputtering under given conditions, the Cr-sputtered surfaces showed superhydrophilic characteristics and its capability became more enhanced with an increase of surface roughness. Judging from spreading behavior of a liquid droplet, the presence of micro-wicking channels, coupled with Cr nanostructures, effectively enhanced the advancing rate of drop base diameter. The CHF exhibited an increasing trend with increasing surface roughness. However, the enhancement ratio agreed poorly with the predictions of the roughness factor-based models, all of which originated from a conventional static force balance.

Thermal hydraulic test for reactor safety system - Critical heat flux experiment and development of prediction models

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; No, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

2000-04-01

To acquire CHF data through the experiments and develop prediction models, research was conducted. Final objectives of research are as follows: 1) Production of tube CHF data for low and middle pressure and mass flux and Flow Boiling Visualization. 2) Modification and suggestion of tube CHF prediction models. 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. The major results of research are as follows: 1) Production of the CHF data for low and middle pressure and mass flux. - Acquisition of CHF data (764) for low and middle pressure and flow conditions - Analysis of CHF trends based on the CHF data - Assessment of existing CHF prediction methods with the CHF data 2) Modification and suggestion of tube CHF prediction models. - Development of a unified CHF model applicable for a wide parametric range - Development of a threshold length correlation - Improvement of CHF look-up table using the threshold length correlation 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. - Development of bundle CHF prediction methodology using correction factor. 11 refs., 134 figs., 25 tabs. (Author)

Critical heat flux measurements in small-diameter tubes using R12 as model fluid

International Nuclear Information System (INIS)

Mueller-Menzel, T.

1987-01-01

Results of critical heat flux measurements are reported for vertical upflow of Refrigerant 12 at high mass fluxes and high pressures in small diameter tubes. The data are transformed into water data using a scaling law, which is verified by means of a new analysis. An error estimation includes the error of the scaling law. Special phenomena ('limiting quality', 'upstream boiling crisis') are explained by theoretical models. The applicability of existing correlations is checked and a new CHF-table for small diameter tubes is presented. With 41 figs., 12 tabs [de

Comparison of CHF predictors obtained by the pseudo-cubic spline method from CEA's FLICA-3M and EDF's THYC computations

International Nuclear Information System (INIS)

Banner, D.; Crecy, F. de.

1993-07-01

Comparison of CHF predictors has been performed by using the same Critical Heat Flux (CHF) databases but by deriving local thermal-hydraulic conditions from different codes (THYC-V3 and FLICA-3M). Predictions have been obtained by the pseudo-cubic Spline method (PCSM). It is shown that the two codes yield similar results provided that equivalent turbulent mixing coefficients are chosen. (author), 6 figs., 5 refs

A study of critical heat flux in the fuel assembly dummies with various types of mixing grids

International Nuclear Information System (INIS)

Bezrukov, Yu. A.; Lisenkov, E. A.; Astakhov, V. I.; Vasilchenko, I. N.

2013-01-01

The report deals with the results of a study The report deals with the results of a study of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m2⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development.of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m 2 ⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development. (authors)

Preliminary Study on CHF Enhancement of Cellulose Nano Fiber (CNF) Fluid with Wire Pool Boiling Experiment

Energy Technology Data Exchange (ETDEWEB)

Hwang, Won Ki; Lee, Yun Seok; Lim, Dong Young; Song, Sub Lee; Lee, Jae Young; Lee, Kwon Yeong [Hanyang Global University, Pohang (Korea, Republic of); Hwang, Dong Soo [POSTECH, Pohang (Korea, Republic of)

2016-05-15

Critical heat flux (CHF) is enhancement of a boiling system will make more compact and effective cooling systems, for examples, nuclear reactors, and air conditioning units. For decades, researchers have been trying to develop more efficient working fluid for heat transfer. This is where nano-fluid could play a key role. There have been a lot of researches for CHF enhancements in nucleate boiling by using nano-fluid which are composed of metal such as copper, Al{sub 2}O{sub 3} and ceramic. And a critical factor of the enhancement is deposition of nano-particles on heating surface, although some results of recent studies are contrary. Also, previous nano-fluid are expensive and have a problem in mass production, so they are difficult to apply to practical industries. Therefore we chose a new material, cellulose nano fiber (CNF) as a solution. CNF can be applied to real situation because it has some advantages which are cost-effectiveness, easiness to get and to make it in nano scale. CHF performance of CNF fluid was different from that of distilled water. Compared to CHF of distilled water, CHF of the CNF fluid which had 0.001V%, 0.01V%, and 0.1V% volumetric concentrations were enhanced to 1%, 104%, and 13% respectively. Likewise other nano-fluid, deposition phenomena was observed in this CNF fluid boiling experiment.

Preliminary Study on CHF Enhancement of Cellulose Nano Fiber (CNF) Fluid with Wire Pool Boiling Experiment

International Nuclear Information System (INIS)

Hwang, Won Ki; Lee, Yun Seok; Lim, Dong Young; Song, Sub Lee; Lee, Jae Young; Lee, Kwon Yeong; Hwang, Dong Soo

2016-01-01

Critical heat flux (CHF) is enhancement of a boiling system will make more compact and effective cooling systems, for examples, nuclear reactors, and air conditioning units. For decades, researchers have been trying to develop more efficient working fluid for heat transfer. This is where nano-fluid could play a key role. There have been a lot of researches for CHF enhancements in nucleate boiling by using nano-fluid which are composed of metal such as copper, Al_2O_3 and ceramic. And a critical factor of the enhancement is deposition of nano-particles on heating surface, although some results of recent studies are contrary. Also, previous nano-fluid are expensive and have a problem in mass production, so they are difficult to apply to practical industries. Therefore we chose a new material, cellulose nano fiber (CNF) as a solution. CNF can be applied to real situation because it has some advantages which are cost-effectiveness, easiness to get and to make it in nano scale. CHF performance of CNF fluid was different from that of distilled water. Compared to CHF of distilled water, CHF of the CNF fluid which had 0.001V%, 0.01V%, and 0.1V% volumetric concentrations were enhanced to 1%, 104%, and 13% respectively. Likewise other nano-fluid, deposition phenomena was observed in this CNF fluid boiling experiment.

Prediction of critical heat flux for water in uniformly heated vertical ...

African Journals Online (AJOL)

Keywords: CHF - Heat transfer - Water vapor - Porous coated tubes. Auteur correspondant ... electrical and mechanical characteristics were well validated. Figure. 1 shows ... resistance to vapor filtration from the heating wall to the liquid bulk.

New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

2016-08-15

Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

A Preliminary Experimental Study on Flow Boiling CHF Characteristics of Ballooned Channel

International Nuclear Information System (INIS)

Kim, Yong Jin; Song, Sub Lee; Chang, Soon Heung; Moon, Sang Ki

2013-01-01

The purpose of this research is to measure heat transfer characteristics experimentally and to develop correlation based on experimental data. Experiments are in progress. The result of preliminary experimental test of ballooned channel was reported. The trends of CHF value for deformed channel is not usual as normal smooth tube. The spot of CHF was moved by changing different experimental cases. The transition of flow pattern at neck of deformation is considered as main factor of changing CHF trends. More cases are under operation and analysis based on flow dynamics are developing. Cladding is one of the most important parts in nuclear power plant because it is second barrier of radiation leakage from nuclear fuel. Originally, cladding keeps its integrity in 1200 .deg. C and 150bar, which is normal operation state of nuclear power plant. However, integrity of cladding can be deformed by more severe conditions caused by accident. In case of LOCA, high temperature, oxidation and thermal shock induced by safety injection can deform cladding. Main problem of deformed cladding is blockage of cooled to prevent core melt accident. Change of flow path by blockage affects flow of safety coolant, heat transfer coefficient and critical heat flux of rod bundles. Until now, there are insufficient heat transfer data for deformed flow path compared to normal flow path. In order to enhance safety of nuclear power plant after accident, it should be clarified that how deformed cladding affects heat transfer

Study on critical heat flux in narrow rectangular channel with repeated-rib roughness. 1. Experimental facility and preliminary experiments

International Nuclear Information System (INIS)

Kinoshita, Hidetaka; Terada, Atsuhiko; Kaminaga, Masanori; Hino, Ryutaro

2001-10-01

In the design of a spallation target system, the water cooling system, for example a proton beam window and a safety hull, is used with narrow channels, in order to remove high heat flux and prevent lowering of system performance by absorption of neutron. And in narrow channel, heat transfer enhancement using 2-D rib is considered for reduction the cost of cooling component and decrease inventory of water in the cooling system, that is, decrease of the amount of irradiated water. But few studies on CHF with rib have been carried out. Experimental and analytical studies with rib-roughened test section, in 10:1 ratio of pitch to height, are being carried out in order to clarify the CHF in rib-roughened channel. This paper presents the review of previous researches on heat transfer in channel with rib roughness, overview of the test facility and the preliminary experimental and analytical results. As a result, wall friction factors were about 3 times as large as that of smooth channel, and heat transfer coefficients are about 2 times as large as that of smooth channel. The obtained CHF was as same as previous mechanistic model by Sudo. (author)

An analysis of critical heat flux on the external surface of the reactor vessel lower head

International Nuclear Information System (INIS)

Yang, Soo Hyung; Baek, Won Pil; Chang, Soon Heung

1999-01-01

CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions

Assessment of 12 CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 computer code

Energy Technology Data Exchange (ETDEWEB)

Ferrouk, M. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria)], E-mail: m_ferrouk@yahoo.fr; Aissani, S. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria); D' Auria, F.; DelNevo, A.; Salah, A. Bousbia [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (Italy)

2008-10-15

The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared.

Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1995-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

International Nuclear Information System (INIS)

Hwang, J. S.; Suh, K. Y.

2009-01-01

The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

Flow visualization and critical heat flux measurement of a boundary layer pool boiling process

International Nuclear Information System (INIS)

Cheung, F.B.; Haddad, K.H.; Liu, Y.C.; Shiah, S.W.

1998-01-01

As part of the effort to evaluate the concept of external passive cooling of core melt by cavity flooding under severe accident conditions, a subscale boundary layer boiling (SBLB) facility, consisting of a pressurized water tank with a condenser unit, a heated hemispherical test vessel, and a data acquisition/photographic system, was developed to simulate the boiling process on the external bottom surface of a fully submerged reactor vessel. Transient quenching and steady-state boiling experiments were conducted in the facility to measure the local critical heat flux (CHF) and observe the underlying mechanisms under well controlled saturated and subcooled conditions. Large elongated vapor slugs were observed in the bottom region of the vessel which gave rise to strong upstream influences in the resulting two-phase liquid-vapor boundary layer flow along the vessel outer surface. The local CHF values deduced from the transient quenching data appeared to be very close to those obtained in the steady-state boiling experiments. Comparison of the SBLB data was made with available 2-D full-scale data and the differences were found to be rather small except in a region near the bottom center of the vessel. The angular position of the vessel outer surface and the degree of subcooling of water had dominant effects on the local critical heat flux. They totally dwarfed the effect of the physical dimensions of the test vessels. (author)

Prediction of the critical heat flux for saturated upward flow boiling water in vertical narrow rectangular channels

International Nuclear Information System (INIS)

Choi, Gil Sik; Chang, Soon Heung; Jeong, Yong Hoon

2016-01-01

A study, on the theoretical method to predict the critical heat flux (CHF) of saturated upward flow boiling water in vertical narrow rectangular channels, has been conducted. For the assessment of this CHF prediction method, 608 experimental data were selected from the previous researches, in which the heated sections were uniformly heated from both wide surfaces under the high pressure condition over 41 bar. For this purpose, representative previous liquid film dryout (LFD) models for circular channels were reviewed by using 6058 points from the KAIST CHF data bank. This shows that it is reasonable to define the initial condition of quality and entrainment fraction at onset of annular flow (OAF) as the transition to annular flow regime and the equilibrium value, respectively, and the prediction error of the LFD model is dependent on the accuracy of the constitutive equations of droplet deposition and entrainment. In the modified Levy model, the CHF data are predicted with standard deviation (SD) of 14.0% and root mean square error (RMSE) of 14.1%. Meanwhile, in the present LFD model, which is based on the constitutive equations developed by Okawa et al., the entire data are calculated with SD of 17.1% and RMSE of 17.3%. Because of its qualitative prediction trend and universal calculation convergence, the present model was finally selected as the best LFD model to predict the CHF for narrow rectangular channels. For the assessment of the present LFD model for narrow rectangular channels, effective 284 data were selected. By using the present LFD model, these data are predicted with RMSE of 22.9% with the dryout criterion of zero-liquid film flow, but RMSE of 18.7% with rivulet formation model. This shows that the prediction error of the present LFD model for narrow rectangular channels is similar with that for circular channels.

Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

International Nuclear Information System (INIS)

Rao, Y.F.; Manzer, A.M.

2005-01-01

This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

Improvement of boiling heat transfer by radiation induced boiling enhancement

International Nuclear Information System (INIS)

Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

2003-01-01

For nuclear reactor systems, the critical heat flux (CHF) data is very important because it limits reactor efficiency. Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, surface wettability in a high-temperature environment and critical heat flux (CHF) of metal oxides irradiated by gamma rays were investigated. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure. The heating test section made of titanium was 0.2 mm in thickness, 3 mm in height, and 60 mm in length. Oxidation of the surface was carried out by plasma jetting for 40 seconds. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. The CHF of oxidized titanium was improved up to 100 percent after 800 kGy 60 Co gamma ray irradiation. We call this effect Radiation Induced Boiling Enhancement (RIBE). Before we conducted the CHF experiment, contact angles of the test pieces were measured to show the relationship between wettability and CHF. The CHF in the present experiment increases will surface wettability in the same manner as shown by Liaw and Dhir's results. (author)

Improvement of boiling heat transfer by radiation induced boiling enhancement

International Nuclear Information System (INIS)

Imai, Y.; Okamoto, K.; Madarame, H.; Takamasa, T.

2003-01-01

For nuclear reactor systems, the Critical Heat Flux (CHF) data is very important because it limits reactor efficiency. Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, surface wettability in a high-temperature environment and Critical Heat Flux (CHF) of metal oxides irradiated by gamma rays were investigated. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure. The heating test section made of titanium was 0.2mm in thickness, 3mm in height, and 60mm in length. Oxidation of the surfaces was carried out by plasma jetting for 40 seconds. The test section was irradiated by 60Co gamma ray with predetermined radiation intensity and period. The CHF of oxidized titanium was improved up to 100 percent after 800kGy 60Co gamma ray irradiation. We call this effect Radiation Induced Boiling Enhancement (RIBE). Before we conducted the CHF experiment, contact angles of the test pieces were measured to show the relationship between wettability and CHF. The CHF in the present experiment increases with surface wettability in the same manner as shown by Liaw and Dhir's results

Nuclear boiling heat transfer and critical heat flux in titanium dioxide-water nanofluids

International Nuclear Information System (INIS)

Okawa, Tomio; Takamura, Masahiro; Kamiya, Takahito

2011-01-01

Nucleate boiling heat transfer was experimentally studied for saturated pool boiling of water-based nanofluids. Since significant nanoparticle deposition on the heated surface was observed after the nucleate boiling in nanofluids, measurement of CHF was also carried out using the nanoparticle deposited heated surface; pure water was used in the CHF measurement. In the present work, the heated surface was a 20 mm diameter cupper surface, and titanium-dioxide was selected as the material of nanoparticles. Experiments were performed for upward- and downward-facing surfaces. Although the CHFs for the downward-facing surface were generally lower than those for the upward-facing surface, the CHFs for the nanoparticle deposited surface were about 1.9 times greater than those for the bare surface in both the configurations. The CHF improvement corresponded well to the reduction of the surface contact angle. During the nucleate boiling in nanofluids, the boiling heat transfer showed peculiar behavior; it was first deteriorated, then improved, and finally approached to an equilibrium state. This observation indicated that the present nanofluid had competing effects to deteriorate and improve the nucleate boiling heat transfer. It was assumed that the wettability and the roughness of the heated surface were influenced by the deposited nanoparticles to cause complex variation of the number of active nucleation sites. During the nucleate boiling of pure water using the downward-facing surface, a sudden increase in the wall temperature was observed stochastically probably due to the accumulation of bubbles beneath the heated surface. Such behavior was not observed when the pure water was replaced by the nanofluid. (author)

Effects of a FeCrAl layer fabricated by sputtering process on pool boiling critical heat flux

International Nuclear Information System (INIS)

Seo, Gwang Hyeok; Son, Hong Hyun; Jeun, Gyoodong; Kim, Sung Joong

2016-01-01

The thermal safety margin of a FeCrAl-layered heater was investigated measuring pool boiling critical heat flux (CHF). Boiling experiments were conducted in a pool of deionized water at atmospheric pressure. For a comparison work, bare and FeCrAl-layered heater samples were prepared. The sputtering technique was employed to fabricate the FeCrAl layer. It was confirmed that the key sputtering parameters on the surface structure were substrate temperature and deposition time. As compared to the bare sample, surface wettability and roughness increased. Higher values of the surface roughness were observed at temperatures of 150degC and 600degC. The FeCrAl-layered heaters showed improved CHF up to ∼40%. The highest enhancement of 42% was observed for the heater sample fabricated at a substrate temperature of 150degC. With employing recent CHF models that incorporate the surface effects, it was evaluated that increased roughness at the micrometer scale mainly contributed to the CHF enhancement. Furthermore, visual observations showed at least 2 msec reduction in the rewetting times for the FeCrAl-layered heaters, and the improved CHF may be attributed to the suppressed hot dry spots due to the rewetting phenomena. (author)

Phenomenological modeling of critical heat flux: The GRAMP code and its validation

International Nuclear Information System (INIS)

Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.

2013-01-01

Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate

Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

External glass peening of zircaloy calandria tubes to increase the critical heat flux

International Nuclear Information System (INIS)

Fong, R.W.L.; Coleman, C.E.; Nitheanandan, T.; Kroeger, V.D.; Moyer, R.G.; Sanderson, D.B.; Root, J.H.; Rogge, R.B.

1997-12-01

Glass-peening the outside surfaces of Zircaloy calandria tubes increases the nucleation sites available for boiling heat transfer and has been demonstrated to enhance the critical heat flux (CHF) in pool-boiling experiments. The objective of this study is to optimise the heat-transfer enhancement by glass peening while ensuring that the microstructure of the peened tube is acceptable for reactor use. Pool-boiling tests were done using small Zircaloy tubes with as-received ('smooth') surfaces and variously peened surfaces, to evaluate two peening parameters, glass-bead size and the coverage of peened surface. Our results showed that the maximum enhancement of CHF (by 60% compared with as-received tubes) was obtained using a glass-bead size of 90-125 μm with a coverage of 100%. The CHF enhancement was found to be insensitive to glass-bead size over a wide range (from 60-90 μm to 125-180 μm). Using a fixed glass-bead size of 125-180 μm to evaluate the influence of peening coverage, the maximum effect on the CHF response was obtained with a coverage of 1 00%. The microstructures of the peened tubes were evaluated using light microscopy, X-ray and neutron diffraction, and mechanical tests. After peening, the microstructure in the subsurface layer (-30 μm) consisted of deformed α-Zr grains, and the crystallographic texture of the grains changed slightly. After stress-relieving at 500 degrees C for 1 h, some recrystallisation had occurred and the residual strains remaining in the tube were low. The tensile and burst properties of glass-peened and stress-relieved tubes were similar to those of as-received tubes. The microstructures introduced by peening and stress relieving were judged to have little effect on creep and growth behaviour. Since there are no deleterious consequences of the glass-peening treatment, the peened and stress-relieved tubes are found to be acceptable for reactor use. (author)

A CHF Model in Narrow Gaps under Saturated Boiling

International Nuclear Information System (INIS)

Park, Suki; Kim, Hyeonil; Park, Cheol

2014-01-01

Many researchers have paid a great attention to the CHF in narrow gaps due to enormous industrial applications. Especially, a great number of researches on the CHF have been carried out in relation to nuclear safety issues such as in-vessel retention for nuclear power plants during a severe accident. Analytical studies to predict the CHF in narrow gaps have been also reported. Yu et al. (2012) developed an analytical model to predict the CHF on downward facing and inclined heaters based on the model of Kandlikar et al. (2001) for an upward facing heater. A new theoretical model is developed to predict the CHF in narrow gaps under saturated pool boiling. This model is applicable when one side of coolant channels or both sides are heated including the effects of heater orientation. The present model is compared with the experimental CHF data obtained in narrow gaps. A new analytical CHF model is proposed to predict CHF for narrow gaps under saturated pool boiling. This model can be applied to one-side or two-sides heating surface and also consider the effects of heater orientation on CHF. The present model is compared with the experimental data obtained in narrow gaps with one heater. The comparisons indicate that the present model shows a good agreement with the experimental CHF data in the horizontal annular tubes. However, it generally under-predicts the experimental data in the narrow rectangular gaps except the data obtained in the gap thickness of 10 mm and the horizontal downward facing heater

A review on critical heat flux in horizontal tubes

International Nuclear Information System (INIS)

Baburajan, P.K.; Gaikwad, Avinash; Prabhu, S.V.

2015-01-01

Coolant channels of PHWR during accident similar to loss of coolant accident (LOCA) may experience different flow transients with low pressure and low flow conditions. In the advanced PHWRs it is desired to have small amount of positive quality at the exit of the coolant channel to increase the thermal efficiency. Investigation on pressure drop and heat transfer coefficient under subcooled boiling condition is important in the design and operation of the PHWRs. Understanding of thermal hydraulic phenomena associated with horizontal flow is also important in the safety and accident management in these reactors. A detailed experimental investigation on the important thermal hydraulic phenomena of horizontal tubes under low pressure and low flow conditions is carried out. The phenomena covered in this work are measurement of diabatic single phase and subcooled boiling pressure drop and local heat transfer coefficients, steady state CHF, effect of upstream flow restrictions on flow transients and CHF, CHF under oscillatory flow and flow decreasing transients. A detailed literature review is carried out on CHF in horizontal channels to take stock of the works being carried out along with current state of the art and to justify the motivation for the experimental study. This paper presents the review of available literature on horizontal CHF with the results of the experimental work. (author)

CHF Enhancement by Surface Patterning based on Hydrodynamic Instability Model

Energy Technology Data Exchange (ETDEWEB)

Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2015-05-15

If the power density of a device exceeds the CHF point, bubbles and vapor films will be covered on the whole heater surface. Because vapor films have much lower heat transfer capabilities compared to the liquid layer, the temperature of the heater surface will increase rapidly, and the device could be damaged due to the heater burnout. Therefore, the prediction and the enhancement of the CHF are essential to maximizing the efficient heat removal region. Numerous studies have been conducted to describe the CHF phenomenon, such as hydrodynamic instability theory, macrolayer dryout theory, hot/dry spot theory, and bubble interaction theory. The hydrodynamic instability model, proposed by Zuber, is the predominant CHF model that Helmholtz instability attributed to the CHF. Zuber assumed that the Rayleigh-Taylor (RT) instability wavelength is related to the Helmholtz wavelength. Lienhard and Dhir proposed a CHF model that Helmholtz instability wavelength is equal to the most dangerous RT wavelength. In addition, they showed the heater size effect using various heater surfaces. Lu et al. proposed a modified hydrodynamic theory that the Helmholtz instability was assumed to be the heater size and the area of the vapor column was used as a fitting factor. The modified hydrodynamic theories were based on the change of Helmholtz wavelength related to the RT instability wavelength. In the present study, the change of the RT instability wavelength, based on the heater surface modification, was conducted to show the CHF enhancement based on the heater surface patterning in a plate pool boiling. Sapphire glass was used as a base heater substrate, and the Pt film was used as a heating source. The patterning surface was based on the change of RT instability wavelength. In the present work the study of the CHF was conducted using bare Pt and patterned heating surfaces.

Experimental study of nonequilibrium post-chf heat transfer in rod bundles

International Nuclear Information System (INIS)

Unal, C.; Tuzla, K.; Badr, O.; Neti, S.; Chen, J.

1986-01-01

Verifications and improvements of nonequilibrium heat transfer models, for post-critical-heat-flux convective boiling, has been greatly affected by the lack of experimental data regarding the degree of thermodynamic nonequilibrium. Recent studies had been successful in measuring vapor superheats in a vertical single tube. This paper extends the nonequilibrium convective boiling data to a rod bundle geometry. Vapor superheat measurements were obtained in a rod bundle with nine heated rods and a heated shroud. Tests were carried out with water at low mass fluxes with a wide range of dryout conditions. Significant nonequilibrium was observed, with vapor superheats of up to 600 0 C. Parametric effects of mass flux, heat flux and inlet conditions on vapor superheat are presented

Post-CHF heat transfer during steady-state and transient conditions

International Nuclear Information System (INIS)

Fung, K.K.

1978-06-01

This review extends previous reviews of steady-state post-CHF literature by Groeneveld, Gardiner, and Fung by including more recent data. A review of the literature on transient post-CHF data is also included by extending the work of Yadigaroglu

Cyro Power and Heat Transfer

National Research Council Canada - National Science Library

Chow, L

1998-01-01

.... The heat generated from a 9x9-heater array was removed by liquid nitrogen pool boiling. The orientation and space limitation of the array were varied to explore their effects on the critical heat flux (CHF) value...

An Experimental Study on the Pool Boiling Heat Transfer on a Square Surface

International Nuclear Information System (INIS)

Kim, Jae Kwang

2000-02-01

An experimental study was carried out to identify the various regimes of natural convective boiling and to determine the Critical Heat Flux (CHF) on a square surface. The basic knowledge on the boiling heat transfer and CHF on the square surface is necessary for various engineering problems, such as the design of compact heat exchangers, cooling of CPU chips, and design of the external cooling mechanism for the reactor during the severe accidents in the nuclear power plants. The heater block made of copper with cartridge heaters in it is submerged in a water tank with windows for visualization. The heater surface has dimension of 70mm x 70mm and the maximum heat flux capacity is about 1.8MW/m 2 . The boiling heat transfer coefficient for the various flow regimes up to CHF has been measured for upward facing surface, vertical surface, and nearly horizontal downward facing surfaces. The temperatures of the heater block are measured by the thermocouples imbedded in the heater block. As the heat flux increases from 100kW/m 2 to 1.0MW/m 2 , the heat-transfer regime changes from the nucleate boiling to the CHF. Near 1.0MW/m 2 , the heat transfer regime suddenly changed from nucleate boiling to film boiling and it resulted in a rapid heat up of the heater block. The various boiling patterns on the vertical surface, upward facing surface, and downward facing surface are observed by a high speed video camera whose frame rate is 1000fps. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux and inclination angle, is observed

Forced convective post CHF heat transfer and quenching

International Nuclear Information System (INIS)

Nelson, R.A.

1980-01-01

This paper discusses mechanisms in the post-CHF region which provide understanding and qualitative prediction capability for several current forced convective heat transfer problems. In the area of nuclear reactor safety, the mechanisms are important in the prediction of fuel rod quenches for the reflood phase, blowdown phase, and possibly some operational transients with dryout. Results using the mechanisms to investigate forced convective quenching are presented. Data reduction of quenching experiments is discussed, and the way in which the quenching transient may affect the results of different types of quenching experiments is investigated. This investigation provides an explanation of how minimum wall superheats greater than the homogeneous nucleation temperature result, as well as how these may appear to be either hydrodynamically or thermodynamically controlled. Finally, the results of a parametric study of the effects of the mechanisms upon the LOFT L2-3 hotpin calculation are presented

Experimental investigation of heat transfer in the transition region

International Nuclear Information System (INIS)

Johannsen, K.; Weber, P.; Feng, Q.

1990-10-01

An experimental study of forced convective boiling heat transfer for upflow of water in a circular tube has been performed using a heat transfer system with temperature-controlled indirect Joule heating. By this way, complete boiling curves from incipience of boiling to fully established film boiling could be measured including the transition boiling regime. The boiling curves were traversed in a quasi-steady mode, usually by increasing the set-point wall temperature average at a constant time rate of 3.5 K/min. The vast majority of results covers the pressure range from 0.1 to 1.0 MPa, mass flux range from 25 to 200 kg/(m 2 s) and inlet subcooling from 5 to 30 K. The experimental results of transition boiling heat transfer obtained in the centre of the test section were correlated in terms of a heat flux/surface superheat relationship that was normalized by the maximum heat flux (local CHF) and its associated wall superheat, respectively, to anchor the transition boiling curve to its low temperature limit. The upper surface temperature limit of the transition boiling regime was determined by inspection of measured axial distributions of surface heat flux and corresponding wall temperature. The critical heat flux (CHF) and its corresponding wall superheat has been measured, too. These temperature-controlled results were compared also with power-controlled experiments. The data are presented in terms of a table and accurate empirical correlations following Katto's generalized correlation scheme. Taking into account previous CHF data at L/D ≤ 100 and same range of flow conditions the length effect was found to further depend on pressure and mass flux. The data for the critical wall superheat show a distinct dependence upon pressure, mass flux and inlet quality that has not been observed before with comparable clarity

An utilization of liquid sublayer dryout mechanism in predicting critical heat flux under low pressure and low velocity conditions in round tubes

International Nuclear Information System (INIS)

Lee, Kwang-Won; Baik, Se-Jin; Ro, Tae-Sun

2000-01-01

From a theoretical assessment of extensive critical heat flux (CHF) data under low pressure and low velocity (LPLV) conditions, it was found out that lots of CHF data would not be well predicted by a normal annular film dryout (AFD) mechanism, although their flow patterns were identified as annular-mist flow. To predict these CHF data, a liquid sublayer dryout (LSD) mechanism has been newly utilized in developing the mechanistic CHF model based on each identified CHF mechanism. This mechanism postulates that the CHF occurrence is caused by dryout of the thin liquid sublayer resulting from the annular film separation or breaking down due to nucleate boiling in annular film or hydrodynamic fluctuation. In principle, this mechanism well supports the experimental evidence of residual film flow rate at the CHF location, which can not be explained by the AFD mechanism. For a comparative assessment of each mechanism, the CHF model based on the LSD mechanism is developed together with that based on the AFD mechanism. The validation of these models is performed on the 1406 CHF data points ranging over P=0.1-2 MPa, G=4-499 kg m -2 s -1 , L/D=4-402. This model validation shows that 1055 and 231 CHF data are predicted within ±30 error bound by the LSD mechanism and the AFD mechanism, respectively. However, some CHF data whose critical qualities are <0.4 or whose tube length-to-diameter ratios are <70 are considerably overestimated by the CHF model based on the LSD mechanism. These overestimations seem to be caused by an inadequate CHF mechanism classification and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities as well as a new bubble detachment model for LPLV conditions, are needed to improve the model accuracy.

CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

International Nuclear Information System (INIS)

Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

2002-10-01

In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

Critical heat flux near the critical pressure in heater rod bundle cooled by R-134A fluid: Effects of unheated rods and spacer grid

International Nuclear Information System (INIS)

Chun, Se-Y.; Shin, C.W.; Hong, S. D.; Moon, S. K.

2007-01-01

A supercritical-pressure light water reactor (SCWR) is currently investigated as the next generation nuclear reactors. The SCWR, which is operated above the thermodynamic critical point of water (647 K, 22.1 MPa), have advantages over conventional light water reactors in terms of thermal efficiency as well as in compactness and simplicity. Many experimental studies have been performed on heat transfer in the boiler tubes of supercritical fossil fire power plants (FPPs). However, the thermal-hydraulic conditions of the SCWR core are different from those of the FPP boiler. In the SCWR core, the heat transfer to the cooling water occurs on the outside surface of fuel rods in rod bundle with spacers. In addition, the experimental studies in which the critical heat flux (CHF) has been carefully measured near the critical pressure have never yet been carried out, as far as we know. Therefore, we have recently conducted the CHF experiments with a vertical 5x5 heater rod bundle cooled by R- 134a fluid. The purpose of this work is to find out some novel knowledge for the CHF near the critical pressure, based on more careful experiments. The outer diameter, heated length and rod pitch of the heater rods are 9.5, 2000 and 12.85 mm, respectively. The critical power has been measured in a range of the pressure of 2.474.03 MPa (the critical pressure of R-134a is 4.059 MPa), the mass flux 502000 kg/m 2 s, and the inlet subcooling 4084 kJ/kg. For the mass fluxes of not less than 550 kg/m 2 s, the critical power decreases monotonously up to the pressure of about 3.63.8 MPa with increasing pressure, and then fall sharply at about 3.83.9 MPa as if the values of the critical power converge on zero at the critical pressure. For the low mass fluxes of 50 to 250 kg/m 2 , the sharp decreasing trend of the critical power near the critical pressure is not observed. The CHF phenomenon near the critical pressure no longer leads to an inordinate increase in the heated wall temperature such as

Two-phase flow characteristic of inverted bubbly, slug, and annular flow in post-critical heat flux region

International Nuclear Information System (INIS)

Ishii, M.; Denten, J.P.

1989-01-01

Inverted annular flow can be visualized as a liquid jet-like core surrounded by a vapor annulus. While many analytical and experimental studies of heat transfer in this regime have been performed, there is very little understanding of the basic hydrodynamics of the post-critical heat flux (CHF) flow field. However, a recent experimental study was done that was able to successfully investigate the effects of various steady-state inlet flow parameters on the post-CHF hydrodynamics of the film boiling of a single phase liquid jet. This study was carried out by means of a visual photographic analysis of an idealized single phase core inverted annular flow initial geometry (single phase liquid jet core surrounded by a coaxial annulus of gas). In order to extend this study, a subsequent flow visualization of an idealized two-phase core inverted annular flow geometry (two-phase central jet core, surrounded by a coaxial annulus of gas) was carried out. The objective of this second experimental study was to investigate the effect of steady-state inlet, pre-CHF two-phase jet core parameters on the hydrodynamics of the post-CHF flow field. In actual film boiling situations, two-phase flows with net positive qualities at the CHF point are encountered. Thus, the focus of the present experimental study was on the inverted bubbly, slug, and annular flow fields in the post dryout film boiling region. Observed post dryout hydrodynamic behavior is reported. A correlation for the axial extent of the transition flow pattern between inverted annular and dispersed droplet flow (the agitated regime) is developed. It is shown to depend strongly on inlet jet core parameters and jet void fraction at the dryout point

Development of the Heated Length Correction Factor

International Nuclear Information System (INIS)

Park, Ho-Young; Kim, Kang-Hoon; Nahm, Kee-Yil; Jung, Yil-Sup; Park, Eung-Jun

2008-01-01

The Critical Heat Flux (CHF) on a nuclear fuel is defined by the function of flow channel geometry and flow condition. According to the selection of the explanatory variable, there are three hypotheses to explain CHF at uniformly heated vertical rod (inlet condition hypothesis, exit condition hypothesis, local condition hypothesis). For inlet condition hypothesis, CHF is characterized by function of system pressure, rod diameter, rod length, mass flow and inlet subcooling. For exit condition hypothesis, exit quality substitutes for inlet subcooling. Generally the heated length effect on CHF in exit condition hypothesis is smaller than that of other variables. Heated length is usually excluded in local condition hypothesis to describe the CHF with only local fluid conditions. Most of commercial plants currently use the empirical CHF correlation based on local condition hypothesis. Empirical CHF correlation is developed by the method of fitting the selected sensitive local variables to CHF test data using the multiple non-linear regression. Because this kind of method can not explain physical meaning, it is difficult to reflect the proper effect of complex geometry. So the recent CHF correlation development strategy of nuclear fuel vendor is making the basic CHF correlation which consists of basic flow variables (local fluid conditions) at first, and then the geometrical correction factors are compensated additionally. Because the functional forms of correction factors are determined from the independent test data which represent the corresponding geometry separately, it can be applied to other CHF correlation directly only with minor coefficient modification

A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

Energy Technology Data Exchange (ETDEWEB)

Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

1995-09-01

Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

Review of Available Data for Validation of Nuresim Two-Phase CFD Software Applied to CHF Investigations

Directory of Open Access Journals (Sweden)

D. Bestion

2009-01-01

Full Text Available The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF. As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis and empirical CHF correlations based on large scale experiments having the real geometry of a reactor assembly. Two-phase CFD is used here for understanding some boiling flow processes, for helping new fuel assembly design, and for developing better CHF predictions in both PWR and BWR. This paper presents a review of experimental data which can be used for validation of the two-phase CFD application to CHF investigations. The phenomenology of DNB and Dry-Out are detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting modeling program of work is given and the current state-of-the-art of the modeling within the NURESIM project is presented.

Occurrence of critical heat flux during blowdown with flow reversal

International Nuclear Information System (INIS)

Leung, J.C.M.

1977-01-01

A small-scale experiment using Freon-11 at 130 0 F (54.4 0 C) and 65 psia (0.45 MPa) in a well-instrumented, transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 61-cm length. Inlet and exit void fractions were measured by a capacitance technique. Flow-regime transition was observed with high-speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F (54.4 0 C) and 60 psig (0.517 MPa) was found to greatly affect the steady-state subcooled-boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the conditions of thermodynamic nonequilibrium required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

A study on critical heat flux in gap

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

1999-04-01

The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs.

A study on critical heat flux in gap

International Nuclear Information System (INIS)

Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

1999-04-01

The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs

CHF enhancement through Pressurized Intermediate Layer in IVR-ERVC Strategy

International Nuclear Information System (INIS)

Park, Seong Dae; Bang, In Cheol

2014-01-01

The molten fuel is sequentially relocated to bottom of reactor vessel. In-vessel retention through the external reactor vessel cooling (IVR-ERVC) strategy has been adapted to some reactors at this situation in order to prevent the progression of an accident. The limitation of IVR-ERVC strategy is CHF phenomenon on the outer wall of reactor vessel. The boiling is main heat transfer mode to remove decay heat between the reactor vessel and the coolant surrounding the reactor vessel. Heated molten radioactive material is leaked. The fuel coolant interaction (FCI) phenomenon could cause the steam explosion in a state of fully flooding condition. Therefore, the CHF should be enhanced in order to be a successful IVR-ERVC strategy. Related studies were performed to confirm the CHF limit with UPLU, SBLB, KAIST and UNIST test facilities The recommendations to increase CHF include coating some materials on the vessel outer surface, increasing the reactor cavity flood level and streamlining the gap between the vessel and the vessel insulation. Recently, flooding the liquid metal is proposed to prevent the boiling itself. In this work, the effects of pressurized liquid layer inserted between the reactor vessel and flooded coolant was studied. Suitable reactor geometry was also presented to apple this concept. Generally, CHF is increased as high pressure was applied until about 1/3 of critical pressure. The limit of IVR-ERVC strategy could overcome by using pressurized intermediate layer. The CFD analysis was performed to confirm the feasibility of pressurized IVR-ERVC system. There are enough thermal margins for due to the enlarged heat transfer area and the convection heat transfer

CHF enhancement through Pressurized Intermediate Layer in IVR-ERVC Strategy

Energy Technology Data Exchange (ETDEWEB)

Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

The molten fuel is sequentially relocated to bottom of reactor vessel. In-vessel retention through the external reactor vessel cooling (IVR-ERVC) strategy has been adapted to some reactors at this situation in order to prevent the progression of an accident. The limitation of IVR-ERVC strategy is CHF phenomenon on the outer wall of reactor vessel. The boiling is main heat transfer mode to remove decay heat between the reactor vessel and the coolant surrounding the reactor vessel. Heated molten radioactive material is leaked. The fuel coolant interaction (FCI) phenomenon could cause the steam explosion in a state of fully flooding condition. Therefore, the CHF should be enhanced in order to be a successful IVR-ERVC strategy. Related studies were performed to confirm the CHF limit with UPLU, SBLB, KAIST and UNIST test facilities The recommendations to increase CHF include coating some materials on the vessel outer surface, increasing the reactor cavity flood level and streamlining the gap between the vessel and the vessel insulation. Recently, flooding the liquid metal is proposed to prevent the boiling itself. In this work, the effects of pressurized liquid layer inserted between the reactor vessel and flooded coolant was studied. Suitable reactor geometry was also presented to apple this concept. Generally, CHF is increased as high pressure was applied until about 1/3 of critical pressure. The limit of IVR-ERVC strategy could overcome by using pressurized intermediate layer. The CFD analysis was performed to confirm the feasibility of pressurized IVR-ERVC system. There are enough thermal margins for due to the enlarged heat transfer area and the convection heat transfer.

An application of liquid sublayer dryout mechanism to the prediction of critical heat flux under low pressure and low velocity conditions in round tubes

International Nuclear Information System (INIS)

Lee, Kwang-Won; Yang, Jae-Young; Baik, Se-Jin

1997-01-01

Based on several experimental evidences for nucleate boiling in annular film and the existence of residual liquid film flow rate at the critical heat flux (CHF) location, the liquid sublayer dryout (LSD) mechanism under annular film is firstly introduced to evaluate the CHF data at low pressure and low velocity (LPLV) conditions, which would not be predicted by a normal annular film dryout (AFD) model. In this study, the CHF occurrence due to annular film separation or breaking down is phenomenologically modelled by applying the LSD mechanism to this situation. In this LSD mechanism, the liquid sublayer thickness, the incoming liquid velocity to the liquid sublayer, and the axial distance from the onset of annular flow to the CHF location are used as the phenomena-controlling parameters. From the model validation on the 1406 CHF data points ranging over P = 0.1 - 2 MPa, G = 4 - 499 kg/m 2 s, L/D = 4 - 402, most of CHF data (more than 1000 points) are predicted within ±30% error bounds by the LSD mechanism. However, some calculation results that critical qualities are less than 0.4 are considerably overestimated by this mechanism. These overpredictions seem to be caused by inadequate CHF mechanism classification criteria and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities and a new bubble detachment model for LPLV conditions are needed to improve the model accuracy. (author)

An improved liquid film model to predict the CHF based on the influence of churn flow

International Nuclear Information System (INIS)

Wang, Ke; Bai, Bofeng; Ma, Weimin

2014-01-01

The critical heat flux (CHF) for boiling crisis is one of the most important parameters in thermal management and safe operation of many engineering systems. Traditionally, the liquid film flow model for “dryout” mechanism shows a good prediction in heated annular two-phase flow. However, a general assumption that the initial entrained fraction at the onset of annular flow shows a lack of reasonable physical interpretation. Since the droplets have great momentum and the length of churn flow is short, the droplets in churn flow show an inevitable effect on the downstream annular flow. To address this, we considered the effect of churn flow and developed the original liquid film flow model in vertical upward flow by suggesting that calculation starts from the onset of churn flow rather than annular flow. The results indicated satisfactory predictions with the experimental data and the developed model provided a better understanding about the effect of flow pattern on the CHF prediction. - Highlights: •The general assumption of initial entrained fraction is unreasonable. •The droplets in churn flow show an inevitable effect on downstream annular flow. •The original liquid film flow model for prediction of CHF was developed. •The integration process was modified to start from the onset of churn flow

Heat flux microsensor measurements

Science.gov (United States)

Terrell, J. P.; Hager, J. M.; Onishi, S.; Diller, T. E.

1992-01-01

A thin-film heat flux sensor has been fabricated on a stainless steel substrate. The thermocouple elements of the heat flux sensor were nickel and nichrome, and the temperature resistance sensor was platinum. The completed heat flux microsensor was calibrated at the AEDC radiation facility. The gage output was linear with heat flux with no apparent temperature effect on sensitivity. The gage was used for heat flux measurements at the NASA Langley Vitiated Air Test Facility. Vitiated air was expanded to Mach 3.0 and hydrogen fuel was injected. Measurements were made on the wall of a diverging duct downstream of the injector during all stages of the hydrogen combustion tests. Because the wall and the gage were not actively cooled, the wall temperature reached over 1000 C (1900 F) during the most severe test.

A study on the development of advanced models to predict the critical heat flux for water and liquid metals

International Nuclear Information System (INIS)

Lee, Yong Bum

1994-02-01

The critical heat flux (CHF) phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal fast breeder reactor (LMFBR). Especially in the LWR application many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed. However, there are few correlations and models which are applicable to liquid metals. Compared with water, liquid metals show a divergent picture for boiling pattern. Therefore, the CHF conditions obtained from investigations with water cannot be applied to liquid metals. In this work a mechanistic model to predict the CHF of water and a correlation for liquid metals are developed. First, a mechanistic model to predict the CHF in flow boiling at low quality was developed based on the liquid sublayer dryout mechanism. In this approach the CHF is assumed to occur when a vapor blanket isolates the liquid sublayer from bulk liquid and then the liquid entering the sublayer falls short of balancing the rate of sublayer dryout by vaporization. Therefore, the vapor blanket velocity is the key parameter. In this work the vapor blanket velocity is theoretically determined based on mass, energy, and momentum balance and finally the mechanistic model to predict the CHF in flow boiling at low quality is developed. The accuracy of the present model is evaluated by comparing model predictions with the experimental data and tabular data of look-up tables. The predictions of the present model agree well with extensive CHF data. In the latter part a correlation to predict the CHF for liquid metals is developed based on the flow excursion mechanism. By using Baroczy two-phase frictional pressure drop correlation and Ledinegg instability criterion, the relationship between the CHF of liquid metals and the principal parameters is derived and finally the

A Review of Wettability Effect on Boiling Heat Transfer Enhancement

International Nuclear Information System (INIS)

Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

2012-01-01

Critical heat flux (CHF) and nucleate boiling heat transfer coefficient (NBHTC) are the key parameters characterizing pool boiling heat transfer. These variables are complicatedly related to thermal-hydraulic parameters of surface wettability, nucleation site density, bubble departure diameter and frequency, to mention a few. In essence, wettability effect on pool boiling heat transfer has been a major fuel to enhance the CHF. Often, however, the improved wettability effect hinders the nucleate boiling. Thus a comprehensive review of such wettability effect may enlighten a further study in this boiling heat transfer area. Phan et al. described surface wettability effects on boiling heat transfer

Implementation of the Westinghouse WRB-2 CHF correlation in VIPRE

International Nuclear Information System (INIS)

Klasmier, L.K.; Haksoo Kim

1992-01-01

As part of the reload transient and thermal-hydraulic methods development effort within Commonwealth Edison Company (CECo), the WRB-2 critical heat flux (CHF) correlation has been implemented into the VIPRE-01 thermal-hydraulic analysis code to support Westinghouse 17X17 Vantage 5 fuel. CECo is in the process of switching from Westinghouse optimized fuel assembly (OFA) fuel to Vantage 5 fuel at CECo's six pressurized water reactors. CECo performs the neutronic portion of the reload analysis using Westinghouse's ANC/PHOENIX. The transient and thermal-hydraulic analysis will be performed using the RETRAN and VIPRE codes once the Nuclear Regulatory Commission has completed their review of CECo methodology. Previously, CECo had implemented and received NRC approval to use the Westinghouse WRB-1 CHF correlation in the VIPRE-01 code to support 15X15 and 17X17 OFA fuel designs. Since the WRB-1 CHF correlation is not applicable for 17X17 Vantage 5 fuel, it was necessary to implement the WRB-2 CHF correlation in the VIPRE code. The WRB-2 correlation was developed by Westinghouse using a database applicable to 17X17 OFA and Vantage 5 fuel and the THINC thermal-hydraulic analysis code. At CECo, the WRB-2 correlation had been implemented into VIPRE-01/MOD-02. The results produced at CECo have been statistically compared to those produced by Westinghouse. Owen's method was used to determine the VIPRE/WRB-02 thermal limit. The thermal limit for 17X17 OFA and Vantage 5 fuel use in VIPRE/WRB-2 is in excellent agreement with the value calculated by Westinghouse using THINC/WRB-2

Assessment of correlations and models for prediction of CHF in subcooled flow boiling

International Nuclear Information System (INIS)

Celata, G.P.; Mariani, A.; Cumo, M.

1992-01-01

This paper provides an analysis of available correlations and models for the prediction of Critical Heat Flux (CHF) in subcooled flow boiling in the ranges of interest of fusion reactor thermal-hydraulic conditions, i.e., high inlet liquid subcooling and velocity and small channel diameter and length. The aim of the study was to establish the limits of validity of present predictive tools (most of them were proposed with reference to LWR thermal-hydraulic studies) in the above conditions. The reference data-set represents most of available data covering wide ranges of operating conditions in the framework of present interest (0.1 s ub, in < 230 K). Among the tens of predictive tools available in literature, four correlations (Levy, Westinghouse, modified-Tong and Tong-75) and three models (Weisman and Ileslamlou Lee and Mudawar and Katto) were selected. The modified-Tong correlation and the Katto model seem to be reliable predictive tools for the calculation of the CHF in subcooled flow boiling

Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

International Nuclear Information System (INIS)

Zoran Stosic; Vladimir Stevanovic

2005-01-01

Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

Application of tube critical heat flux tables to annuli and rod bundles

International Nuclear Information System (INIS)

Ulrych, G.

1985-01-01

The purpose of this paper is to show that tables for the critical heat flux (CHF) in tubes have a much wider range of applicability than only to tubes. With the proper choice of a characteristic length replacing the tube diameter as a parameter the validity of the tables can be expanded to more complex geometries. The paper describes how the tables must be applied to annuli or rod bundles. The data base for comparisons is mainly taken from the open literature. For rod bundles the proposed methodology was checked for very different geometries including rod bundles from very tight hexagonal to extremely open square bundle arrays. It is concluded that the tables give reasonable results for a wide range of hydraulic diameters

Exploring the Limits of Boiling and Evaporative Heat Transfer Using Micro/Nano Structures

OpenAIRE

Lu, Ming-Chang

2010-01-01

This dissertation presents a study exploring the limits of phase-change heat transfer with the aim of enhancing critical heat flux (CHF) in pool boiling and enhancing thermal conductance in heat pipes. The state-of-the-art values of the CHF in pool boiling and the thermal conductance in heat pipes are about two orders of magnitudes smaller than the limits predicted by kinetic theory. Consequently, there seems to be plenty of room for improvement. Pool boiling refers to boiling at a surface im...

Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow

International Nuclear Information System (INIS)

Boscary, J.

1995-10-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs

Subcooled flow boiling heat transfer of dilute alumina, zinc oxide, and diamond nanofluids at atmospheric pressure

International Nuclear Information System (INIS)

Kim, Sung Joong; McKrell, Tom; Buongiorno, Jacopo; Hu Linwen

2010-01-01

A nanofluid is a colloidal suspension of nano-scale particles in water, or other base fluids. Previous pool boiling studies have shown that nanofluids can improve the critical heat flux (CHF) by as much as 200%. In a previous paper, we reported on subcooled flow boiling CHF experiments with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (≤0.1% by volume) at atmospheric pressure, which revealed a substantial CHF enhancement (∼40-50%) at the highest mass flux (G = 2500 kg/m 2 s) and concentration (0.1 vol.%) for all nanoparticle materials (). In this paper, we focus on the flow boiling heat transfer coefficient data collected in the same tests. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient are similar (within ±20%). The heat transfer coefficient increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. A confocal microscopy-based examination of the test section revealed that nanoparticle deposition on the boiling surface occurred during nanofluid boiling. Such deposition changes the number of micro-cavities on the surface, but also changes the surface wettability. A simple model was used to estimate the ensuing nucleation site density changes, but no definitive correlation between the nucleation site density and the heat transfer coefficient data could be found.

The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies

International Nuclear Information System (INIS)

De Crecy, F.

1994-01-01

Critical heat flux (CHF) is one of the limiting phenomena for a PWR. It has been widely studied for years, but many facts are still not satisfactorily understood. This paper deals with the effect of the grid assembly mixing vanes on both the value of the CHF and the azimuthal location of the departure from nucleate boiling (DNB). A series of experimental studies was performed on electrically heated, 5x5 square pitched, vertical rod bundles. Two specific grid assembly designs were used: with and without mixing vanes. DNB was detected by eight thermocouples welded internally in each rod at the same level in order to determine the azimuthal location. The coolant was Freon-12 flowing upwards to simulate high pressure water (as defined by Stevens). Single-phase flow experiments were also conducted to measure the exit temperature field in order to obtain the mixing coefficients for subchannel analysis.The results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value. - Without mixing vanes, DNB occurs mainly on the most central rod and preferentially at the azimuthal location facing the adjacent rod. - With mixing vanes, DNB can occur on any of the nine central rods and is distributed in an apparently random way around the rod. -The effect of the mixing vanes on CHF is dramatic and depends a great deal on the parameter range (pressure, local mass velocity and local quality). Generally speaking, CHF with mixing vanes is significantly higher than without mixing vanes, but this effect can be inverted in some cases.In order to understand this fact more clearly, it is necessary to perform detailed analysis of subchannel behavior. Indeed, the analyses show that the magnitude of this effect is closely related to the mixing coefficients used. These mixing coefficients, estimated from the single-phase flow experiments, are subject to large uncertainties in two-phase flow. ((orig.))

Review of Critical Heat Flux Correlations for Upward Flow in a Vertical Thin Rectangular Channel

International Nuclear Information System (INIS)

Choi, Gil Sik; Chang, Soon Heung

2014-01-01

From the view point of safety, this type of fuel has higher resistance to earthquake and external impact. The cross section of coolant flow channel in the reactor core composed with the plate fuel is a thin rectangular shape. Thermal-hydraulic characteristics of this thin rectangular channel are different with those of general circular rod fuel bundle flow channel. Accordingly it could be thought that the CHF correlation in a thin rectangular channel is different with that in a circular channel, for which a large number of researches on CHF prediction have been carried out. The objective of this paper is to review previous researches on CHF in a thin rectangular channel, summarize the important conclusion and propose the new simple CHF correlation, which is based on the data set under high pressure and high flow rate condition. The researches on CHF in rectangular channel have been partially carried out according to the pressure, heated surface number, heated surface wettability effect, flow driving force and flow direction conditions. From the literature researches on CHF for upward flow in a vertical thin rectangular channel, some CHF prediction methods were reviewed and compared. There is no universal correlation which can predict CHF at all conditions, but generally, Katto empirical correlation is known to be useful at high pressure and high flow rate. The new simple correlation was developed from the restricted data set, the CHF prediction capacity of which is better than that of Katto. Even though the prediction consistency of the new simple correlation is lower, MAE and RMS error decreased quite. For the more development of the new simple CHF correlation, the more advanced regression analysis method and theoretical analysis should be studied in future

Review of Critical Heat Flux Correlations for Upward Flow in a Vertical Thin Rectangular Channel

Energy Technology Data Exchange (ETDEWEB)

Choi, Gil Sik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

From the view point of safety, this type of fuel has higher resistance to earthquake and external impact. The cross section of coolant flow channel in the reactor core composed with the plate fuel is a thin rectangular shape. Thermal-hydraulic characteristics of this thin rectangular channel are different with those of general circular rod fuel bundle flow channel. Accordingly it could be thought that the CHF correlation in a thin rectangular channel is different with that in a circular channel, for which a large number of researches on CHF prediction have been carried out. The objective of this paper is to review previous researches on CHF in a thin rectangular channel, summarize the important conclusion and propose the new simple CHF correlation, which is based on the data set under high pressure and high flow rate condition. The researches on CHF in rectangular channel have been partially carried out according to the pressure, heated surface number, heated surface wettability effect, flow driving force and flow direction conditions. From the literature researches on CHF for upward flow in a vertical thin rectangular channel, some CHF prediction methods were reviewed and compared. There is no universal correlation which can predict CHF at all conditions, but generally, Katto empirical correlation is known to be useful at high pressure and high flow rate. The new simple correlation was developed from the restricted data set, the CHF prediction capacity of which is better than that of Katto. Even though the prediction consistency of the new simple correlation is lower, MAE and RMS error decreased quite. For the more development of the new simple CHF correlation, the more advanced regression analysis method and theoretical analysis should be studied in future.

A study on the flow induced vibration in two phase flow under heating and non-heating conditions

International Nuclear Information System (INIS)

Kim, Dae Hun

2007-02-01

Critical heat flux (CHF) enhancement devices, like a spacer grid with mixing vane, cause flow-induced vibration (FIV) due to turbulence made by structural resistance. CHF enhancement and FIV reduction are usually studied separately. The main purpose of this article is to investigate the relationship between CHF and FIV. Information of flow-induced vibration due to wire coil design, is experimentally presented in this study by detecting flow-induced vibration under the two-phase flow condition with wire coil inserts. CHF experiments were performed in an upward vertical annulus tube under controlled vibration conditions to determine the effect of vibration on CHF. FIV was measured in an upward vertical tube with various wire coil inserts using air-water as flow material. CHF experiments were performed at one atmosphere with mechanically controlled vibration. A quartz tube (inner diameter of 17 mm, thickness of 2mm and length of 0.72 m) was used for outer tube and a SUS-304 tube (outer diameter of 6.35 mm, thickness of 0.89 mm and length of 0.7 m) was used for the inner heater. Vibration of the heater tube with an amplitude range of 0.1 mm to 0.5 mm and a frequency range of 10 Hz to 50 Hz was carried out at a mass flux of 115 kg/m 2 s and 215 kg/m 2 s. CHF was enhanced by vibration with a maximum ratio of 16.4 %. CHF was increased with increased amplitude and quality. The CHF correlation was developed with R (coefficient of correlation) of 0.903. FIV measuring experiments were performed at one atmosphere by changing the inserted wire coil type. An acrylic tube was used for the test section with inner diameter of 25 mm, thickness of 10 mm and length of 0.5 m. Four types of wire coil, which have a thickness of between 2 mm and 3 mm and pitch length of between 25 mm and 50 mm, were used. FIV and dynamic pressure were detected in water mass flux range of 100 ∼ 3060 kg/m 2 s and air mass flux range of 5.02 ∼ 60.3 kg/m 2 s. Vibration increased along with mass flux and

Dry-out heat fluxes of falling film and low-mass flux upward-flow in heated tubes

International Nuclear Information System (INIS)

Koizumi, Yasuo; Ueda, Tatsuhiro; Matsuo, Teruyuki; Miyota, Yukio

1998-01-01

Dry-out heat fluxes were investigated experimentally for a film flow falling down on the inner surface of vertical heated-tubes and for a low mass flux forced-upward flow in the tubes using R 113. This work followed the study on those for a two-phase natural circulation system. For the falling film boiling, flow state observation tests were also performed, where dry-patches appearing and disappearing repeatedly were observed near the exit end of the heated section at the dry-out heat flux conditions. Relation between the dry-out heat flux and the liquid film flow rate is analyzed. The dry-out heat fluxes of the low mass flux upflow are expressed well by the correlation proposed in the previous work. The relation for the falling film boiling shows a similar trend to that for the upflow boiling, however, the dry-out heat fluxes of the falling film are much lower, approximately one third, than those of the upward flow. (author)

Prediction of liquid film dryout in two-phase annular-mist flow in a uniformly heated narrow tube development of analytical method under BWR conditions

International Nuclear Information System (INIS)

Utsuno, Hideaki; Kaminaga, Fumito

1998-01-01

A method was developed based on the conservation lows to predict critical heat flux (CHF) causing liquid film dryout in two-phase annular-mist flow in a uniformly heated narrow tube under BWR conditions. The applicable range of the method is within the pressure of 3-9 MPa, mass flux of 500-2,000 kg/m 2 ·s, heat flux of 0.33-2.0 MW/m 2 and boiling length-to-tube diameter ratio of 200-800. The two-phase annular-mist flow was modeled with the three-fluid streams with liquid film, entrained droplets and gas flow. Governing equations of the method are mass continuity and energy conservation on the three-fluid streams. Constitutive equations on the mass transfer which consist of the entrainment fraction at equilibrium and the mass transfer coefficient were newly proposed in this study. Confirmation of the present method were performed in comparison with the available film flow measurements and various CHF data from experiments in uniformly heated narrow tubes under high pressure steam-water conditions. In the heat flux range (q'' 2 ) practical for a BWR, agreement of the present method with CHF data was obtained as, (Averaged ratio) ± (Standard deviation) = 0.984 ± 0.077, which was shown to be the same or better agreement than the widely-used CHF correlations. (author)

Evaluation of CHF experimental data for non-square lattice 7-rod bundles

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; Kim, K. K.; Zee, S. Q

2001-01-01

A series of CHF experiments are conducted for 7-rod hexagonal test bundles in order to investigate the CHF characteristics of self-sustained square finned (SSF) rod bundles. The experiments are performed in the freon-loop and water-loop located at IPPE in Russia, and 609 data of freon-12 and 229 data of water are obtained from 7 kinds of test bundles classified by the combination of heated length and axial/radial power distributions. As the result of the evaluation of four representative CHF correlations, the EPRI-1 correlation reveals good prediction capability for SSF test bundles. The inlet parameter CHF correlation suggested by IPPE calculates the mean and the standard deviation of P/M for uniformly heated test bundles as 1.002 and 0.049, respectively. In spite of its excellent accuracy, the correlation has a discontinuity point at the boundary between the low velocity and high velocity conditions. KAERI's inlet parameter correlation eliminates this defect by introducing the complete evaporation model at low velocity condition, and calculates the mean and standard deviation of P/M as 0.095 and 0.062 for uniformly heated 496 data points, respectively. The mean/standard deviation of local parameter CHF correlations suggested by IPPE and KAERI are evaluated as 1.023/0.178 and 1.002/0.158, respectively. The inlet parameter correlation developed from uniformly heated test bundles tends to under-predict CHF about 3% for axially non-uniformly heated test bundles. On the other hand, the local parameter correlation reveals large scattering of P/M, and requires re-optimization of the correlation for non-uniform axial power distributions. As the result of the analysis of experimental data, it reveals that the correction model of axial power shapes suggested by IPPE is applicable to the inlet parameter correlations. For the test bundle of radial non-uniform power distribution, the physically unexpected results are obtained at some experimental conditions. In addition

A study on the upward and downward facing pool boiling heat transfer characteristics of graphene-modified surface

International Nuclear Information System (INIS)

Kim, Ji Hoon; Ahn, Ho Seon; Kim, Ji Min

2016-01-01

Recently, graphene, carbon in two dimensions, were highlighted as a good heat transfer materials, according to its high thermal conductivity. Lateral conduction and water absorption into the structure helped graphene films to inhibit the formation of hot spots, which means increasing of critical heat flux (CHF) and boiling heat transfer coefficient (BHTC) performances. In this study, we report a promising increase of CHF and BHTC results with 2D graphene films. Furthermore, we tried to observe bubble behavior via high-speed visualization to investigate a relationship between bubble behavior and pool boiling performances in downward facing boiling. The effect of graphene film coating on the pool boiling performances of upward and downward facing heater surface were examined. 2D- and 3D- graphene film showed good enhancement results on the CHF (by 111% and 60%) and BHTC (by 40% and 20-25%) performances. Bubble behavior change was significant factor on the CHF and BHTC performances in downward facing boiling. The amount of evaporation heat flux was calculated from the velocity, bubble diameter, frequency, orientation angle and superheat that the post-products of the high-speed visualization

Flow visualization study of post critical heat flux region for inverted bubbly, slug and annular flow regimes

International Nuclear Information System (INIS)

Denten, J.G.; Ishii, M.

1988-11-01

A visual study of film boiling using still photographic and high- speed motion picture methods was carried out in order to analyze the post-CHF hydrodynamics for steady-state inlet pre-CHF two-phase flow regimes. Pre-CHF two-phase flow regimes were established by introducing Freon 113 liquid and nitrogen gas into a jet core injection nozzle. An idealized, post-CHF two-phase core initial flow geometry (cylindrical multiphase jet core surrounded by a coaxial annulus of gas) was established at the nozzle exit by introducing nitrogen gas into the annular gap between the jet nozzle two-phase effluent and the heated test section inlet. For the present study three basic post-CHF flow regimes have been observed: the rough wavy regime (inverted annular flow preliminary break down), the agitated regime (transition between inverted annular and dispersed droplet flow), and the dispersed ligament/droplet regime. For pre-CHF bubbly flow in the jet nozzle, the post-CHF flow (beginning from jet nozzle exit/heated test section inlet) consists of the rough wavy regime, followed by the agitated and then the dispersed ligament/droplet regime. In the same way, for pre-CHF slug flow in the jet core, the post-CHF flow is comprised of the agitated regime at the nozzle exit, followed by the dispersed regime. Pre-CHF annular jet core flow results in a small, depleted post-CHF agitated flow regime at the nozzle exit, immediately followed by the dispersed ligament/droplet regime. Observed post dryout hydrodynamic behavior is reported, with particular attention given to the transition flow pattern between inverted annular and dispersed droplet flow. 43 refs., 20 figs., 5 tabs

Analysis of heat transfer mechanism on in-vessel corium coolability in severe accidents

International Nuclear Information System (INIS)

Park, Rae Joon; Jeong, Ji Whan; Kim, Sang Baik; Kang, Kyung Ho; Kim, Jong Whan

1998-04-01

When the molten core material relocates to the lower plenum of the reactor vessel, the cooling process of corium and the related heat transfer mechanism have been analyzed. The critical heat flux in gap (CHFG) test is being performed as a part of simulation of naturally arrested thermal attack in (SONATA-IV) project and the state of art on CHF has been reviewed. A series of complex heat transfer mechanism of molten pool formation, natural convection in the molten pool, solidification and remelting of the corium, conduction in the solidified crust, and boiling heat transfer to surroundings can be occurred in the lower plenum. Many studies are needed to investigate the complex heat transfer mechanism in the lower plenum, because these phenomena have not been clearly understand until now. The SONATA-IV/CHFG experiments are being carried out to develop CHF correlation in a hemispherical gap, which is the upper limit of heat transfer. There is no experimental or analytical CHF correlation applicable to a hemispherical gap. So lots of analytical and experimental correlations developed using the similar experimental condition were gathered and compared with each other. According to the experimental work that was carried out with pool boiling condition, CHF in a parallel gap was reduced by 1/30 compared with the value measured without gap. A basic form of a CHF correlation has been developed to correlate measurements that will be made in the SONATA-IV/CHFG experiments. That correlation is based on the fact that the CHF in a hemispherical gap is enhanced by CCFL and a Kutateladze type CCFL correlation develops CCFL date will in geometry like this. The experimental facility consists of a heater, a pressure vessel, a heat exchanger and lots of sensors. The heater capacity is 40 kw and the maximum heat flux at the surface is 100 kw/m 2 . The experiments will be carried out in the range of 1 to 10 atm and the gap size of 0.5, 1, 2 mm. The CHF will be detected using 66 type

Study of Two-Phase Heat Transfer in Nano-fluids for Nuclear Applications

International Nuclear Information System (INIS)

Kim, S.J.; Truong, B.; Buongiorno, J.; Hu, L.W.; Bang, I.C.

2006-01-01

Nano-fluids are engineered colloidal suspensions of nano-particles in a base fluid. We are investigating the two-phase heat transfer behavior of water-based nano-fluids, to evaluate their potential use in nuclear applications, including the PWR primary coolant and PWR and BWR safety systems. A simple pool boiling wire experiment shows that a significant increase in Critical Heat Flux (CHF) can be achieved at modest nano-particle concentrations. For example, the CHF increases by 50% in nano-fluids with alumina nano-particles at 0.001%v concentration. The CHF enhancement appears to correlate with the presence of a layer of nano-particles that builds up on the heated surface during nucleate boiling. A review of the prevalent Departure from Nucleate Boiling (DNB) theories suggests that an alteration of the nucleation site density (brought about by the nano-particle layer) could plausibly explain the CHF enhancement. (authors)

Application of ANNS in tube CHF prediction: effect on neuron number in hidden layer

International Nuclear Information System (INIS)

Han, L.; Shan, J.; Zhang, B.

2004-01-01

Prediction of the Critical Heat Flux (CHF) for upward flow of water in uniformly heated vertical round tube is studied with Artificial Neuron Networks (ANNs) method utilizing different neuron number in hidden layers. This study is based on thermal equilibrium conditions. The neuron number in hidden layers is chosen to vary from 5 to 30 with the step of 5. The effect due to the variety of the neuron number in hidden layers is analyzed. The analysis shows that the neuron number in hidden layers should be appropriate, too less will affect the prediction accuracy and too much may result in abnormal parametric trends. It is concluded that the appropriate neuron number in two hidden layers should be [15 15]. (authors)

Sensors for Metering Heat Flux Area Density and Metrological Equipment for the Heat Flux Density Measurement

Science.gov (United States)

Doronin, D. O.

2018-04-01

The demand in measuring and studies of heat conduction of various media is very urgent now. This article considers the problem of heat conduction monitoring and measurement in various media and materials in any industries and branches of science as well as metrological support of the heat flux measurement equipment. The main study objects are both the sensors manufactured and facilities onto which these sensors will be installed: different cladding structures of the buildings, awnings, rocket fairings, boiler units, internal combustion engines. The Company develops and manufactures different types of heat flux sensors: thermocouple, thin-film, heterogeneous gradient as well as metrological equipment for the gauging calibration of the heat flux density measurement. The calibration shall be performed using both referencing method in the unit and by fixed setting of the heat flux in the unit. To manufacture heterogeneous heat flux gradient sensors (HHFGS) the Company developed and designed a number of units: diffusion welding unit, HHFGS cutting unit. Rather good quality HHFGS prototypes were obtained. At this stage the factory tests on the equipment for the heat flux density measurement equipment are planned. A high-sensitivity heat flux sensor was produced, now it is tested at the Construction Physics Research Institute (Moscow). It became possible to create thin-film heat flux sensors with the sensitivity not worse than that of the sensors manufactured by Captec Company (France). The Company has sufficient premises to supply the market with a wide range of sensors, to master new sensor manufacture technologies which will enable their application range.

A study on the CHF enhancement of pool boiling using nano-fluids

International Nuclear Information System (INIS)

Chang, Won Joon

2009-02-01

The understanding of CHF phenomenon and an accurate prediction of the CHF condition are important for safe and economic design of many heat transfer units including nuclear reactors, fossil fuel boilers, fusion reactors, electronic chips, etc. The phenomenon has been investigated extensively over the world since Nukiyama (1934) first characterized it. In particular, a large amount of significant work has been done during the last four decades with the development of water cooled nuclear reactors. The wettability of the heated surface under pool boiling of surfactant solutions and nano-fluids has been investigated. Tri-sodium phosphate (TSP, Na 3 PO 4 ) solutions and Aluminum oxide nano-fluids were prepared for experiments. Contact angles of pure water and the solutions on the quenched surface and fresh surface were measured. Surfaces deposited TSP and nano-particle could affect surface energy of the strips and enhance hydrophilicity of the surfaces. Several implications of the experimental results on the pool boiling CHF model and CHF enhancement using TSP and NF were discussed. A increase of CHF was observed with nano-fluid. The addition of nano-particle helped to increase the wettability by reducing the surface tension. This happens with the decrease in bubble diameter, breakup of bubbles and avoidance of bubble coalescence. CHF increase or decrease depends upon competition between high wettability and high instability. An optimum nano-fluid concentration is needed which must have high crystalline content. When the concentration reaches at a critical value, CHF will tend to a constant value. As the results of previous study, surface tension effect the results of CHF. And it is same to nano-fluids, because surface tension change the dynamics of mixture fluids at two phase and means the instability of thermal hydraulics. Contact angle which be in the limelight at recent research means wettability of heated surface. However, in case of nano-fluids, both are

Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Sugandhi, Ritesh, E-mail: ritesh@ipr.res.in; Swamy, Rajamannar, E-mail: rajamannar@ipr.res.in; Khirwadkar, Samir, E-mail: sameer@ipr.res.in

2016-11-15

Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

International Nuclear Information System (INIS)

Sugandhi, Ritesh; Swamy, Rajamannar; Khirwadkar, Samir

2016-01-01

Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

An empirical correlation for the entrainment fraction at the onset of annular flow based on 2006 CHF look-up table

Energy Technology Data Exchange (ETDEWEB)

Jiao, Bo; Yang, Dongyu [Department of Mechanical Engineering, Rongcheng Campus, Harbin University of Science and Technology, Rongcheng 264300, Shandong (China); Gan, Zhihua, E-mail: gan_zhihua@zju.edu.cn [Institute of Refrigeration and Cryogenics, Zhejiang University, Hangzhou (China); Key Laboratory of Refrigeration and Cryogenic Technology of Zhejiang Province, Hangzhou (China); National Quality Inspection Center of Refrigeration Equipment (Henan), Minquan (China)

2017-06-15

Highlights: • A parallel look-up table for the entrainment fraction at annular point was developed. • A correlation was given based on the selected database from 2006 CHF look-up table. • Its reliability on the other conditions in the look-up table was discussed. - Abstract: The critical heat flux (CHF) of gas-liquid flow plays an important role in the safety of industrial equipment. At present, the liquid film dryout model is widely used for predicting CHF in gas-liquid annular flow. Most parameters in this model can be determined by some empirical correlations which are valid under different conditions. However, up to now, the entrainment fraction at the onset of annular flow is always assumed due to the lack of relevant experimental data. In this paper, the normalized data of the 2006 CHF look-up table (LUT) which has been adopted widely, especially in the nuclear industry, were used. Firstly, the empirical correlations, provided for the onset of annular flow and the limiting quality, were employed. In the valid pressure and mass flux range of these correlations, the selected database from LUT was confirmed. Secondly, the liquid film model was built. The entrainment fraction at the onset of annular flow was obtained when the calculated CHF by the model agreed with the corresponding value in LUT. A parallel look-up table for it was developed. Its correlation including the Weber and the liquid Reynolds number at outlet was proposed. The errors are mostly within ±30%. Finally, its reliability on the other conditions in LUT, which are beyond the valid range of the empirical correlations used for determining the database, was discussed. All the conditions whose errors are outside ±30% of the predictions by the provided correlation were marked in the tables.

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

International Nuclear Information System (INIS)

Buongiorno, Jacopo; Hu, Lin-wen

2009-01-01

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 ((le) 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/m 2 s) 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 used to estimate the ensuing nucleation site

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

Energy Technology Data Exchange (ETDEWEB)

Buongiorno, Jacopo; Hu, Lin-wen

2009-07-31

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 used to estimate the ensuing nucleation site

FILM-30: A Heat Transfer Properties Code for Water Coolant

International Nuclear Information System (INIS)

MARSHALL, THERON D.

2001-01-01

A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating

Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

OpenAIRE

H. P. Rahardjo; V. I. Sri Wardhani

2017-01-01

The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

Measurement of pool boiling CHF for SUS 304 and SA 508 flat plate under downward-facing and atmospheric conditions

International Nuclear Information System (INIS)

Kam, Dong Hoon; Park, Hae Min; Choi, Young Jae; Jeong, Yong Hoon

2015-01-01

Heat transfer performance of downward-facing conditions are important especially in severe accident mitigation strategy (IVR-ERVC and Core-catcher). Heat transfer limit, in other word, critical heat flux (CHF) is important value in this basis to guarantee the integrity of the system. For the application point of view in nuclear power plant, carbon steel surface should also be considered since reactor pressure vessel (RPV) in IVR-ERVC strategy consists of carbon steel, and core-catcher in EU-APR1400 is also composed of carbon steel. In this perspective, carbon steel surface was used in previous studies. In this study, CHF of both stainless steel and carbon steel material were measured under pool boiling condition with various inclination angles and dimensions. There was a width effect as angle increases, but it disappeared as approached to horizontally downward condition. Besides, there was almost no length effect for both of the width since the size of coalesced bubble was far smaller than the length of short test section (100 mm). SA 508 showed enhanced results at high angles for 40 mm-width case even though no oxidation occurred on the surface during the experiments

Measurement of pool boiling CHF for SUS 304 and SA 508 flat plate under downward-facing and atmospheric conditions

Energy Technology Data Exchange (ETDEWEB)

Kam, Dong Hoon; Park, Hae Min; Choi, Young Jae; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

2015-05-15

Heat transfer performance of downward-facing conditions are important especially in severe accident mitigation strategy (IVR-ERVC and Core-catcher). Heat transfer limit, in other word, critical heat flux (CHF) is important value in this basis to guarantee the integrity of the system. For the application point of view in nuclear power plant, carbon steel surface should also be considered since reactor pressure vessel (RPV) in IVR-ERVC strategy consists of carbon steel, and core-catcher in EU-APR1400 is also composed of carbon steel. In this perspective, carbon steel surface was used in previous studies. In this study, CHF of both stainless steel and carbon steel material were measured under pool boiling condition with various inclination angles and dimensions. There was a width effect as angle increases, but it disappeared as approached to horizontally downward condition. Besides, there was almost no length effect for both of the width since the size of coalesced bubble was far smaller than the length of short test section (100 mm). SA 508 showed enhanced results at high angles for 40 mm-width case even though no oxidation occurred on the surface during the experiments.

Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube

International Nuclear Information System (INIS)

Boscary, J.; Association Euratom-CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

1997-03-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author)

A study on the effect of the CHF correlations to the LOCA analysis

International Nuclear Information System (INIS)

Kim, Ho Kee

1998-02-01

The critical heat flux (CHF) is a major parameter which determines the cooling performance and therefore the prediction of CHF is of importance for the design and safety analysis in boiling systems; such as nuclear reactors, conventional boilers, and other various two-phase flow systems. Until now, many CHF correlations have been developed and for the actual design a correlation has been selected in consideration of its characteristics. For the analysis of Loss of Coolant Accident (LOCA) in a Nuclear Power Plant, which shows the drastic parameters change during the system transient, a correlation having a reasonable degree of accuracy over a wide range is preferred, rather than that having accuracy for a specific range. It is required to have tangible insight about the effects of the CHF correlation to the LOCA analysis for the purpose of computer code development and nuclear regulation. The related research is further recommended. The purpose of this research is to obtain an insight and/or intuition about the above effect and to evaluate the selected CHF correlations. To achieve these purposes LOCA is analysed for the UL-JIN 3 and 4 nuclear power plant, the Korea Standard Type Nuclear Power Plant and the Loss of Flow Test (LOFT) L2-5 experiment is simulated using the RELAP5/MOD3.1 computer code for each selected CHF correlation. The selected correlations are the AECL-UO Lookup Table, adapted in RELAP5 code; the K110 CHF correlation, developed by KAERI; and the original W-3 CHF correlation, developed by L.S. Tong. LOFT is also simulated using the AECL-UO Lookup Table having the CHF multiplication factors 0.5 and 1.5, and then compared with the result of the original Lookup Table and the experiment result. In the LOCA analysis, the CHF correlations affect the magnitude of peak cladding temperatures, but does not seriously affect the occurrence points of time. The effect of each CHF correlation to the fuel cladding temperature behavior becomes apparent at the end of

Establishment and assessment of CHF data base for square-lattice rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

2002-02-01

A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

CHF predictor derived from a 3D thermal-hydraulic code and an advanced statistical method

International Nuclear Information System (INIS)

Banner, D.; Aubry, S.

2004-01-01

A rod bundle CHF predictor has been determined by using a 3D code (THYC) to compute local thermal-hydraulic conditions at the boiling crisis location. These local parameters have been correlated to the critical heat flux by using an advanced statistical method based on spline functions. The main characteristics of the predictor are presented in conjunction with a detailed analysis of predictions (P/M ratio) in order to prove that the usual safety methodology can be applied with such a predictor. A thermal-hydraulic design criterion is obtained (1.13) and the predictor is compared with the WRB-1 correlation. (author)