Experimental Studies on Heat Transfer by Natural Convection and Pool Boiling of Sodium in a Strong Magnetic Field

This report deals with an experiment on the heat transfer of liquid sodium, with particular reference to the effects brought by the application of a magnetic field on pool boiling. The test section, a heater pin of 6.5 mm diameter, was inserted vertically into the center of a sodium tank. The heating surface of the pin was parallel to the magnetic field as well as to the direction of gravity.

Under conditions of natural convection in a magnetic field, a sharp rise of the heating surface temperature was always seen to occur at some point when the heat flux was gradually increased, accompanied by the onset of sharp temperature oscillations.

The surface superheat required for the initiation of boiling decreased with increasing intensity of the applied magnetic field, reached a minimum, then increased again.

The surface temperature fluctuations in nucleate boiling was higher under magnetic field than when free of such influence. The critical heat flux for burn-out was not appreciably affected by magnetic field.     

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

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/m² s) and concentration (0.1 vol.%) for all nanoparticle materials (Kim et al., 2009). 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.     

Critical heat flux of forced convection boiling in an oscillating acceleration field — III. Reduction mechanism of CHF in subcooled flow boiling

To investigate the effect of variation in acceleration on the critical heat flux (CHF) in subcooled flow boiling, a photographic study was made. The test section was an internally heated vertical annulus with a glass shroud, in which Freon-113 flowed upwardly. The observation was made at a pressure of 3 bar, a mass flux of 920 kg/m²s, an inlet subcooling 45 K and a slightly lower heat flux level than steady CHF. The vertical acceleration was oscillated with amplitude of 0.3g_e and a period of 6 s.At low apparent gravitational acceleration, bubbles generated on the heated surface moved longer along the surface without detachment and coalesced with other bubbles to form large vapor slugs. This causes early CHF, the mechanism of which is dry-out of the liquid film existing between the heated surface and vapor slugs.     

Critical heat flux in a vertical tube at low and medium pressures. Part II — new data representation

An analysis of the physical processes taking place in a dispersed-annular flow which govern dry-out type CHFs has been carried out. The analysis has shown that the number of variables required to describe the critical phenomena can be reduced by the introduction of a new parameter: the length over which dispersed-annular flow takes place, L_dan. In this case only, for a given tube diameter, pressure and mass flux, the critical heat flux may be expressed in terms of a single variable: L_dan. A correlation which may be used to determine this length has also been developed. The representation of the CHF data obtained at low pressures in terms of the coordinate system (L_dan, q″_cr) has shown that the dispersion of the data about the regression curves is considerably reduced as compared with the traditional presentation of the critical heat flux as a function of the thermodynamic quality at the end of the heated length.     

Method of Critical Power Prediction Based on Film Flow Model Coupled with Subchannel Analysis

A new method was developed to predict critical powers for a wide variety of BWR fuel bundle designs. This method couples subchannel analysis with a liquid film flow model, instead of taking the conventional way which couples subchannel analysis with critical heat flux correlations. Flow and quality distributions in a bundle are estimated by the subchannel analysis. Using these distributions, film flow rates along fuel rods are then calculated with the film flow model. Dryout is assumed to occur where one of the film flows disappears. This method is expected to give much better adaptability to variations in geometry, heat flux, flow rate and quality distributions than the conventional methods.

In order to verify the method, critical power data under BWR conditions were analyzed. Measured and calculated critical powers agreed to within ±7%. Furthermore critical power data for a tight-latticed bundle obtained by LeTourneau et al. were compared with critical powers calculated by the present method and two conventional methods, CISE correlation and subchannel analysis coupled with the CISE correlation. It was confirmed that the present method can predict critical powers more accurately than the conventional methods.     

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention Part I: Observation of quenching of downward-facing surfaces

This paper presents the results of a series of quenching experiments to examine the boiling curve of relative large downward-facing surfaces. The test masses are 61 cm in diameter and the downward-facing surface is either flat or curved. The work is motivated by the need to assess the ex-vessel boiling process for in-vessel core retention. The critical heat flux is found to be approximately 50 W cm⁻². The nucleate boiling regime and the critical heat flux regime are found to be characterized by cyclic two-phase flow patterns.     

Prediction of Liquid Film Dryout in Two-Phase Annular-Mist Flow in a Uniformly Heated Narrow Tube Development of Analytical Method under BWR Conditions

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/m²·s, heat flux of 0.33–2.0 MW/m² 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“<2MW/m²) 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.     

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract

The behavior of neutrons in a highly heterogeneous unit cell consisting of D₂O moderator, H₂O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different ²³⁵U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D₂O level, as well as various lattice parameters

A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H₂O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H₂O coolant.

The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D₂O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p ²⁸—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.     

Heat Transfer to Steam-Water Annular Dispersed Flow in Vertical Annulus

The heat transfer coefficient and slow burnout heat flux were measured for a stream-water annular dispersed upward flow under pressures up to 3.5 ata in an electrically heated vertical annular channel.

An empirical equation was derived for the heat transfer coefficient as function of mass flow rate, steam quality and heat flux. The dominant mechanism of heat transfer to the annular dispersed two-phase flow is forced convection of liquid film on the heater surface even in the region of low steam quality (down to about 0.03). The observed slow burnout heat flux was near the point of intersection of the lines representing liquid film forced convective heat transfer and nucleate boiling heat transfer on the q vs. δT _sat diagram. A dryout mechanism is proposed in which increasingly violent evaporation comes to impede the rewetting of the dry patches generated on the heater surface, which thus spread to cover the whole surface. A maximum value is observed in the slow burnout heat flux plotted against exit steam quality. This can be explained as the effect of heat removal by droplet exchange between liquid film and steam flow.     

Design and thermal performance of an improved mechanically attached module for divertor plate of LHD

A mechanically attached divertor module with improved performance has been developed and partially installed in the large helical device (LHD). The advantage of the new design is to eliminate any metal bolts on the armor tile surface which leads to high Z impurity emission. The new module consists of a couple of armor tiles made of iso-graphite and a thin graphite sheet, which are tightened by two stainless steel (SS) or TZM bolts horizontally sandwiching a SS cooling pipe. The heat flows directly from the tile surface to the cooling pipe. The previous design used a copper heat sink has been used. Steady heat flux tests up to 1.2 MW/m² have been carried out for the new module without any trouble using a electron gun ACT while the allowable heat flux of the previous module was limited below 0.3 MW/m² due to deformation of the copper heat sink. The outgassing from the new module during heat flux tests up to 0.5 MW/m² is decreased to about one-third. Thermal fatigue test up to 115 cycles under a steady heat flux of 1 MW/m² for the new module has been performed without any troubles. Comparisons between the previous and new modules for the structure, thermal performance, and outgassing are made.     

Heat Transfer and Critical Heat Flux in Transient Boiling, (I)

Heat transfer and critical heat flux in saturated pool boiling were experimentally studied under transient power condition. The heating elements were flat plates of nickel submerged facing vertically in stagnant water. The heat generation rate in the test section was increased linearly in time, upon which, under certain conditions, the heat flux was found to reach a maximum point located in the nucleate boiling regime. The heat flux of this critical point increased with mounting sharpness of the transient, and the mechanism that occurs such a high critical heat flux may be the rapid formation and evaporation of thin liquid film at the bases of vapor bubbles. Examination of high speed motion pictures reveals that all bubbles on the heating surface are still in the phase of the first generation until the critical condition is reached. Compared to the case of steady boiling, the effect of differences in the heat capacity of the test section upon the critical heat flux was found to be less marked under the present experimental conditions.     

Direct Measurement of Neutron Flux

An experimental study was carried out on steady-state boiling of sodium flowing in the annular channel formed around an electrically heated simulation of a fuel pin. In the present experiment, the inlet temperature and flow rate were held constant, and the heat flux was gradually increased up to the inception of boiling. Thereafter, the heat flux was further increased step by step until the surface temperature of the heater pin marked a sharp rise, indicating the occurrence of dry-out. Records were obtained of the changes brought by the increasing heat flux to boiling phenomena, with particular reference to the behavior of the two-phase flow pattern and to the characteristics of boiling noise, as well as of the frequency of bubble formation.

It was made clear that there exists a region in which steady-state boiling will be established, and under these conditions the two-phase flow pattern changes sequentially from bubbly flow to slug flow and then to annular flow. This behavior of sodium boiling in a narrow channel is quite similar to the case of water.

With rising heat flux, the level of noise intensity associated with boiling first increased sharply to attain a maximum point, then decreased somewhat and remained constant thereafter until dry-out.

The frequency of bubble formation depended on the size of the bubble. The product of bubble frequency and equivalent diameter was found to be constant.     

A subcooled boiling heat transfer predictive model for ITER EHF FW designs

Recent experimental data from the ITER critical heat flux (CHF) mock-ups was used to benchmark a 3D CFD code concerning subcooled boiling heat transfer for high heat flux removal. The predicted temperatures show good agreement with experimental measurements for a range of operating parameters and of cooling configurations. Specifically, it applies to a hypervapotron channel exposed to a 5 MW/m² surface heat load and cooled by velocity of 2 m/s. Such flow geometry and operating condition seem necessary for ITER-enhanced heat flux first wall modules if an adequate design margin in CHF is needed. A detailed CFD and heat transfer analysis performed on a prototyped CAD model provided a higher confidence on the design and is deemed a desirable feature for continued design exploration and optimization processes. This is particularly crucial in regard to flow distribution among the FW fingers.     

A new model for predicting the critical heat flux based on nucleation site density in downward-face boiling

ABSTRACT

Due to the important role critical heat flux (CHF) plays in the boiling field, it is of great significance to study CHF, especially the mechanism of CHF in the nucleate boiling. In this study, a new model to predict CHF both in pool boiling and flow boiling of downward-face was proposed and the relationship between CHF and nucleation site density (NSD) was studied. The model was based on the bubble interaction theory, which assumed that CHF happened due to the coalescing of the bubbles generated on the heating surface and prevented liquid to be supplied. The relationship between NSD and CHF was derived from previous observations in the experiments and simulations. To validate the relationship between NSD and CHF, several experiments with CHF and NSD were chosen and they all showed good agreement with our assumptions. Due to the rarity of experimental data on NSD and CHF, the numerical method was also used to validate. The results also showed an inverse relationship between CHF and NSD.     

Analysis of the conjugation effect during convective flow film boiling conditions in an eccentric annular test section

The objective of this investigation was to present a technique for estimating the conjugation effect on post-dryout heat transfer. To this aim, an experimental study was conducted to measure the conjugation effect on the post-dryout heat transfer using an internally heated eccentric annular test section (outside diameter (O.D.), 13.51 mm; inside diameter (I.D.), 11.54 mm). The experiments were carried out at 7 and 9 MPa with the mass flux varying from 2.0 to 5.0 Mg m⁻² s⁻¹ and the inlet vapour quality from 0.01 to 0.20. Five different minimum gap sizes between 0.06 and 1.92 mm were tested. The two-dimensional well temperature distribution on the inside surface of the heated tube was measured using a unique sliding thermocouple technique. A data reduction method was developed to determine the radial heat flux variation on the boiling surface from the temperature measurements. A numerical smoothing procedure was used to minimize the effect of the random temperature measurement error of the sliding thermocouple on the estimated radial heat flux variation. The results show that the conjugation effect can cause the local radial heat flux to deviate by as much as 17% from the average value.     

Prediction of critical heat flux in vertical pipe flow

A previously developed semi-empirical model for adiabatic two-phase annular flow is 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.     

Studies on cooling by two-dimensional confined jet flow of high heat heat flux surface in fusion reactor

Divertor surface of a magnetic confinement fusion reactor is exposed to strong radiative heating. According to standard design of the ITER, maximum heat flux on the divertor surface becomes locally near 30 MW m⁻². To cool such high heat flux surface by water flow, it is necessary to establish a cooling method which enhanced the critical heat flux (CHF). We proposed a cooling by a planar impinging jet with free surface in the previous report. In the jet cooling on flat surface, high CHF was obtained in the limited region where the jet flow hits directly. As apart from the region, the CHF decreases abruptly with the distance from the center. To overcome this difficulty, it was proposed that the planar jet is applied to cool concave surface where the centrifugal force is efficiently used to enhance the CHF. In this study, the CHFs were investigated in the confined jet flow which was guarded by a wall on the other side of the heated wall, because the guard wall works to protect splash of water from liquid film by violent boiling and expects further enhancement of the CHF. In this study, the CHFs were investigated in the confined flow of two-dimensional jet on flat and concave surfaces in the various flow conditions and got a correlation for the CHF. Applicability of this cooling for divertor surface was assessed by using the experimental results.     

超亲水特性传热面上水喷流冷却时的沸腾换热特性实验研究

研究了在纳米材料涂层形成的超亲水材料传热表面上水喷流冷却时的沸腾换热特性,并与普通金属面上的喷流沸腾特性进行了比较.介绍了流动条件、加热方式、加热条件、超亲水材料涂层等条件对高温沸腾换热、临界热流密度的定量影响和一些基本换热规律.实验发现,超亲水传热表面对沸腾换热特性有十分显著的影响,超亲水传热面强化了喷流沸腾临界热流密度.     

Measurement and Evaluation on Pulsing Characteristics and Experimental Capability of NSRR

The NSRR programme is in progress in JAERI using a pulsed reactor to investigate fuel behaviors under the reactivity-initiated accident conditions. Pulsing characteristics and experimental capability, especially heat deposition in test fuel rods given by a single pulse are key parameters to this purpose.

In pulsing performance tests, it has been ascertained that the maximum pulsing with 4.67$ (=3.41%δk) brings peak reactor power of 21,100 MW and core energy release of 117 MW·sec. The calculated time responses of reactor power, fuel temperature and cladding surface temperature as well as these maximum values at various pulse sizes agreed well with measured data. In addition, it has been also ascertained by measurement as well as analysis that there are no essential differences in pulsing characteristics between the pulsing from critical and that from subcritical.

The heat deposition in a test fuel rod given by a single pulse is much enough as predicted, and a 2.6% enriched BWR type fuel rod gains about 230cal/g-UO₂ in the maximum pulsing. In case of irradiation of clustered five test fuel rods by a single pulse, heat deposition reduces by about 20% for a surrounding rod and about 40% for a center rod in comparison with that in a single rod irradiation.     

A simulated plasma disruption experiment using an electron beam as a heat source

An experimental study was made on the behavior of a solid surface subjected to an extremely high heat flux similar to that expected during a plasma disruption. An electron beam was used as the heat source to simulate the high heat flux. The beam was defocused in an attempt to give as much uniform heat flux as possible on the test surface. The 5-mm-diameter test pieces were made of 304 stainless steel, aluminum, and zinc. Heat fluxes from 10 to 110 MW/m² were applied on the test pieces for durations of 90 to 180 msec. Special attention was paid to the measurement of the surface heat flux on the test surface. Comparison between experimental and analytical results on melt layer thickness and evaporation loss is made. An improved thermal analysis code (DAT-K) was developed for the analysis. Agreement between the experimental and analytical results on melt layer thickness is good. For evaporation loss, experimental and analytical results are in fair agreement. Features of the experiments and analysis that lead to the differences in the results are discussed.On leave from Kawasaki Heavy Industries, Ltd.