Heater Rod Research Articles

Exergy, economic and environmental analysis of forced circulation flat plate solar collector using heat transfer enhancer in riser tube

This paper experimentally investigates the exergy of a riser tube with heat transfer enhancer in the flat plate collector. Two different internal heat transfer enhancers are used herein, namely, rod heat transfer enhancer and tube heat transfer enhancer. The environmental and sustainable assessment of the flat plate collector with and without heat transfer enhancer have been taken into consideration in the study. This is the first study performed on the exergy and economic analysis of a flat plate solar collector using this heat transfer enhancer, and the new findings on the exergy analysis using heat transfer enhancer of the flat plate collector have been reported. The improved exergy efficiency, useful exergy rate and exergy destruction for the flat plate collector are analyzed, and it is found that the rod heat transfer enhancer provides comparatively higher exergy efficiency than the tube heat transfer enhancer and plain tube flat plate collector. The maximum exergy efficiency is 11.3%, 10.9% and 8.3% for the rod, tube heat transfer enhancers, and plain tube respectively. The heat transfer enhanced collectors significantly reduce the total exergy destruction compared to plain tube collector. The increase in useful exergy rate is 1.29 and 1.25 times higher for the rod and tube heat transfer enhancers respectively, compared to the plain tube collector. The conclusion is that the rod heat transfer enhancer collector has the lowest energy payback time and a simple payback period.

Validated Simulations of Heat Transfer From a Vertical Heated-Rod Array to a Helium-Filled Isothermal Enclosure

Measurements of heat transfer from an array of vertical heater rods to the walls of a square, helium-filled enclosure are performed for a range of enclosure temperatures, helium pressures, and rod heat generation rates. This configuration is relevant to a used nuclear fuel assembly within a dry storage canister. The measurements are used to assess the accuracy of computational fluid dynamics (CFD)/radiation simulations in the same configuration. The simulations employ the measured enclosure temperatures as boundary conditions and predict the temperature difference between the rods and enclosure. These temperature differences are as large as 72 °C for some experiments. The measured temperature of rods near the periphery of the array is sensitive to small, uncontrolled variations in their location. As a result, those temperatures are not as useful for validating the simulations as measurements from rods near the array center. The simulated rod temperatures exhibit random differences from the measurements that are as large as 5.7 °C, but the systematic (average) error is 1 °C or less. The random difference between the simulated and measured maximum array temperature is 2.1 °C, which is less than 3% of the maximum rod-to-wall temperature difference.

Experimental investigation of void fraction variation in subcooled boiling flow under horizontal forced vibrations

An experimental investigation of horizontal forced vibration effect on void fraction variation of subcooled boiling flow was carried out in this study. In order to simulate the fuel assembly subchannel of a boiling water reactor (BWR), an annular test section with inner and our diameters of 19.1 and 38.1mm was utilized for subcooled boiling tests under atmospheric pressure. The annular test section was attached to an eccentric-cam vibrator, which was driven by a low-speed motor and can produce horizontal forced vibrations with frequency up to 20Hz and maximum displacement of 22.2mm. The inlet liquid velocity and subcooling were set as vf,in=0.25–1.00m/s and ΔTSub=5–20°C. Different heat fluxes of q″=0.058–0.193MW/m2 were loaded through the center heater rod, and the void fraction and fluid temperature were measured during the tests under stationary (no vibration) and vibration conditions. Test results show that in the subcooled boiling region, the void fraction and fluid temperature can vary under horizontal forced vibrations, and the variation trends were presented in NZu-NSub and vf,in-〈α〉 plots. These variations can be explained by the potential changes of thermal boundary layers (TBL) and the heat transfer enhancement under vibrations. In addition, no significant change of void fraction and fluid temperature was found in or near the saturated boiling conditions under vibrations.

Experimental investigation on transient heat transfer in 2 × 2 bundle during depressurization from supercritical pressure

The experimental investigation on transient heat transfer of supercritical water during depressurization in 2×2 bundle with wire wraps has been performed on the SWAMUP-II co-constructed by CGNPC and SJTU. The test section consists of two channels separated by a square steel assembly box with round corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2×2 bundle with wire wraps installed inside the assembly box. The bundle consists of four heater rods of 10mm in O.D. arranged with pitch-to-diameter ratio of 1.18. Experimental parameter ranges cover pressure from 16 to 26MPa, mass flux from 850 to 1450kg/m2s, heat flux from 250 to 650kW/m2, inlet fluid temperature from 345 to 365°C, and depressurization rate of 1 and 2MPa/min. The experimental data are obtained and heat transfer characteristics during depressurization from supercritical to sub-critical conditions are discussed. The fraction of heat transfer from inner channel to outer channel is lower than 15% under the test conditions. Four different transient heat transfer phenomena during depressurization are observed. The boiling crisis is likely to occur in SCWR during depressurization as the operating condition of SCWR is compared with the test condition. The boiling crisis is less likely to occur with the increase of mass flux, and more likely to occur with the increase of heat flux or inlet fluid temperature. The wall temperature increment during boiling crisis is smaller with higher mass flux, and larger with higher heat flux or inlet fluid temperature. The depressurization rate affects the transient heat transfer itself trivially.

Experimental study on the thermal stratification in a pool boiling with a horizontal heat source

Large pools of water near atmospheric pressure have been incorporated into several advanced nuclear reactor designs. These pools provide a heat sink to remove heat from the reactor or containment by natural circulation and are a source of water for core cooling. Thermal stratification is formed in horizontal fluid layers with different temperatures, where the warmer fluid layers are situated above the cooler fluid layers. In the present study, the key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The two-dimensional velocity vector fields that occur as the water temperature increases to the saturation temperature are experimentally investigated in a pool that contains a horizontal heater rod. The detailed velocity measurements using the 2D PIV measurement technique were conducted to investigate single- and/or two-phase natural convection flow and thermal stratification in a pool boiling. The experimental results indicate that a large natural convection flow occurs above the heater rod and that thermal stratification occurs below the heater rod. The flow in the opposite directions to each other was shown in the region between the heater rod and the thermal boundary layer. This flow pattern will contribute to maintain the thermal stratification and retard the water temperature rise and the dissipation of the thermal boundary layer in a pool boiling with a horizontal heat source installed asymmetrically. The current experimental data can be directly used to develop and validate turbulence models in a pool boiling. Furthermore, the present experimental results can help characterize the performance of boiling models on heated surfaces and provide benchmark data to validate the calculation performance of a component scale analysis code and CFD code.

Subcooled boiling-induced vibration of a heater rod located between two metallic walls

The phenomenon that a heating structure vibrates itself due to the behavior of vapor bubbles generated under subcooled boiling has been known as “Subcooled Boiling-induced Vibration (SBIV)”. As one of such a heating structure, fuel assemblies for Boiling Water Reactors (BWR) are utilized in subcooled boiling of water, and those for Pressurized Water Reactors (PWR) may face unexpected subcooled boiling conditions in case of sudden drop of the system pressure or loss of water flow, though they are utilized in single phase of water under normal operating conditions. As studies on SBIV, some researchers have conducted demonstrative experiments with a partial array of fuel rods simulating the actual BWR fuel assembly in a flow test loop, which showed no significant influences of the SBIV to degrade the integrity of the fuel rods. In addition, in order to investigate the fundamental phenomenon of the SBIV, pool boiling experiments of the SBIV on a single heater rod were performed in other studies with a simplified apparatus of a water tank in laboratory size under atmospheric pressure. In the experiments, behavior of bubbles generated under various degree of subcooling were observed, and the acceleration of the SBIV of the heater rod was measured. The present study, as a series of the above experiments for the fundamental phenomenon of the SBIV, the two thin walls made of stainless steel were installed in parallel to interleave the heater rod with the gap distance of 1mm or 3mm to each of the two walls, which was expected to provide useful results to investigate effects of the parallel walls configuration to the SBIV of the heater rod. The results obtained in the experiments suggested that bubbles growing largely in the gap influenced on the SBIV of the heater rod, with the remarkable behavior such as slide of the large bubbles on the surface of the heater rod towards the open region, interference and coalescence between large bubbles.

Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all the models. We find that boiling nonlinearity and void-dependent power removal from the primary loop increases the supercritical region while inclusion of clad temperature to include fuel rod heat transfer increases the subcritical regime.

Experimental investigation on heat transfer and pumping power of forced circulation flat plate solar collector using heat transfer enhancer in absorber tube

This paper experimentally investigates the performance of an absorber tube with heat transfer enhancer in solar flat plate water heater. The objective of the work is to increase the convective heat transfer by reducing the cross sectional area between the absorbing fluid and inner wall of the tube. The heat transfer enhancers are frictionally engaged with the inner side of the tube wall, and it is kept in the axial flow direction of the fluid flow path. Two types of heat transfer enhancers, namely, rod heat transfer enhancer and tube heat transfer enhancer, are used herein and compared with each other. The efficiency, heat transfer coefficient and pumping power for flat plate solar collector were analyzed, and it was found that, the rod heat transfer enhancer provides higher heat transfer with a small increase in pumping power than tube heat transfer enhancer and plain tube flat plate solar collector. The maximum increase in pumping power is 1.081 and 1.044 times higher for rod and tube heat transfer enhancers respectively, compared to plain tube solar collector.

Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

Abstract A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 × 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

가열탱크 내부 스크류 회전속도에 따른 전기보일러의 열성능에 관한 연구

Abstract This study was aimed at the heating tank with a screw-rotation device for improving the thermalefficiency of electric boiler. In the proposed system, analysis items were the heater rod surfacetemperature variation, reaching time for set temperature and thermal efficiency. The followingconclusions are obtained from this experimental study. (1) When screw speed increases, the timereaching for set temperature tended to be shorter. (2) When the rotation speed becomes 300 rpm, thesurface temperature difference between the right and left heater rod decreases by 49%, from 19.7 o C to9.7 o C in average. (3) When the rotation speed is over 250 rpm, proposed heating tank structure appearedto be effective in terms of thermal efficiency. Thermal efficiency with the rotation speed 300 rpm isimproved by 3.8% compared to the case of rotation speed 0 rpm.Keywords: Electric boiler(전기보일러 ), Heater rod(히터봉 ), Heating tank(가열탱크 ), Screw(스크류 ),rpm(회전수 ), Thermal Efficiency(열효율 ) 기호설명 G : 유량 [kg/s]C : 유체의 비열 [kJ/kg

The Effect Of Carbon Concentration On The Retained Austenite Content And The Mechanical Properties Of TRIP Steel Wire Rod Obtained From The Stelmor Controlled Cooling Line

Abstract The austenite content of the multiphase TRIP-structure steels depends, inter alia, on the carbon concentration and the properly selected parameters of the two-stage heat treatment. Under the existing industrial conditions, it is possible to (approximately) reproduce approximately the optimal parameters of the two-stage wire rod heat treatment via the controlled wire rod cooling from the end temperature of rolling on the Stelmor line. The investigation of the retained austenite content of TRIP wire rods with a varying carbon concentration, produced under industrial conditions, has been discussed and the effect of the multiphase structure of these wire rods on their mechanical properties has been determined in the paper.

Experimental study on heat transfer to supercritical water in 2 × 2 rod bundle with wire wraps

Experimental studies on heat transfer to supercritical water in 2×2 rod bundle with wire wraps are performed at Shanghai Jiaotong University. The test section consists of two channels separated by a square steel assembly box with round corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2×2 rod bundle with wire wraps installed inside the assembly box. The bundle consists of four heater rods of 10mm in O.D. and 1.18in. pitch-to-diameter ratio. Experimental parameter ranges cover pressure from 23.0MPa to 26.0MPa, mass flux from 400kg/m2s to 1400kg/m2s, heat flux from 300kW/m2 to 1000kW/m2 and bulk temperature from 280°C to 500°C. The fluid enthalpy in the first channel increases due to the heat transfer through the assembly box when flowing downward. Heat transfer deterioration is observed in the bundle. The degree of the deterioration is moderated in the developed region due to the mixing by the wire wraps. Significant non-uniformity of circumferential wall-temperature distribution around the heater rods is observed. Effects of various parameters on heat transfer behavior inside the 2×2 rod bundle are similar to those observed in a tube. The heat transfer enhancement due to the wire wraps becomes remarkable under a high mass flux condition. The Jackson and Fewster correlation and Bishop et al. correlation give the best predictions when evaluated against the experimental data.

Experimental Investigation of Inverted Annular Film Boiling in a Rod Bundle during Reflood Transient

Heat transfer results for subcooled and saturated inverted annular film boiling (IAFB) obtained from a 7×7 rod bundle during transient reflood are presented in this paper. The test section consists of heater rods of 9.5-mm diameter and 12.6-mm pitch arranged in a square array. Flooding rates considered are 0.076 and 0.152 m/s, pressure varied from 138 to 414 kPa, and inlet subcooling up to 83 K. Evaluation of the data includes estimation of the local void fraction and Nusselt number during IAFB as well as in the inverted slug film boiling (ISFB) regime, which occurs when the inverted annular liquid column disintegrates. Experimental heat transfer results are compared with several film boiling models, and a new correlation for the Nusselt number is proposed for the IAFB and ISFB regimes. Predicted Nusselt numbers using the new correlation deviate from the experimental data by an average error of 15% and root-mean-square error of ∼30%.

Study of Critical Heat Flux in Natural Convection–Cooled TRIGA Reactors with Single Annulus and Rod Bundle Geometries

A critical heat flux (CHF) experimental study at low pressure and natural convection condition has been conducted. The test apparatus is a natural circulation loop with an upward flow channel, simulating TRIGA (Training, Research, Isotopes, General Atomics) reactors. CHF is studied in three types of geometries: a single-rod annulus, a three-rod bundle in a trefoil tube, and a four-rod bundle in a square tube. The full-scale fuel pin heater rod is electrically heated with a prototypic axial power profile, equipped with thermocouples for CHF detection. Experiments are carried out at the following conditions: inlet subcooling from 10 to 70 K, pressure from 110 to 290 kPa, and mass flux from 0 to 400 kg/m2·s. It is observed that CHF increases as the pressure or mass flux increases but does not significantly depend on the inlet subcooling within the testing range. The current CHF data are compared with a few selected CHF correlations whose application ranges are close to the testing conditions. The relevance of the CHF to the testing parameters is investigated. A modified CHF correlation compatible with TRIGA reactor conditions is proposed based on a previous correlation and current experimental data.

Experimental and numerical investigation of heat transfer and pressure drop for innovative gas cooled systems

Heat transfer enhancement through turbulence augmentation is recognized as a key factor for improving the safety and economic conditions in the development of both critical and subcritical innovative advanced gas cooled fast reactors (GFR) and transmutation systems. The L-STAR facility has been designed and erected at the Karlsruhe Institute of Technology (KIT) to study turbulent flow behavior and its heat transfer enhancement characteristics in gas cooled annular channels under a wide range of conditions. The test section consists of an annular hexagonal cross section channel with an inner electrical heater rod element, placed concentrically within the test section, which seeks to simulate the flow area of a fuel rod element in a GFR. The long term objective of the experimental study is to investigate and improve the understanding of complex turbulent convective enhancement mechanisms as well as the friction loss penalties of roughened fuel rods compared to smooth ones and to generate an accurate database for further development of physical models. In the first step, experimental results of the fluid flow with uniform heat release conditions for the smooth heater rod are presented. The pressure drops, as well as the axial temperature profiles along the heater rod surface have been measured at Reynolds numbers in the range from 4000 to 35,000. The experimental results of the first stage were compared with independently conducted CFD analyses performed at Lappeenranta University of Technology (LUT) with the code ANSYS Fluent and at Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) with the code ANSYS CFX.

Experimental investigation on the natural convection flow in pool boiling

In the present study, the key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The volumetric scaling ratio of the test rig is 1/910 the size of the passive auxiliary feedwater system (PAFS) condensing heat removal assessment loop (PASCAL), which is a PAFS performance evaluation test facility. The two-dimensional velocity vector fields that occur as the water level decreases are experimentally investigated in a pool that contains a horizontal heater rod. The 2D particle image velocimetry (PIV) measurement technique is adopted to determine the velocity vector field of the natural convection flow. The experimental results indicate that a large natural convection flow occurs above the heater rod and that thermal stratification occurs below the heater rod. The thermal stratification and the stagnant region begin to disappear when the pool temperature reaches approximately 90°C. The experimental results can provide benchmark data to validate computational fluid dynamics (CFD) calculations of thermal hydraulic phenomena that occur in a pool with a heat source.

Development of fine-grain size titanium 6Al–4V alloy sheet material for low temperature superplastic forming

Fine-grained titanium 6Al–4V alloy, which typically has a grain size of about 1–2μm, can be made to superplastic form at around 800°C with special processing. The normal temperature for superplastic forming (SPF) with conventional titanium 6Al–4V sheet material is 900°C. The lower temperature performance is of interest to the Boeing Company because it can be exploited to achieve significant cost savings in processing by reducing the high-temperature oxidation of the SPF dies, improving the heater rod life for the hot presses, increasing operator safety and replacing the chemical milling operation to remove alpha case contamination with a less intensive nitric hydrofluoric acid etchant (pickle). In this report, room temperature tensile tests and elevated temperature constant strain rate tensile tests of fine-grained Ti–6Al–4V sheets provided by the Baoti Company of Xi׳an, China, were conducted according to the test method standards of ASTM-E8 and ASTM-E2448. The relationships among the processing parameters, microstructure and superplastic behavior have been analyzed. The results show that two of the samples produced met the Boeing minimum requirements for low-temperature superplasticity. The successful material was heat-treated at 800°C subsequent to hot rolling above the beta transus temperature, Tβ-(150–250°C). It was found that the sheet metal microstructure has a significant influence on superplastic formability of the Ti–6Al–4V alloy. Specifically, fine grains, a narrow grain size distribution, low grain aspect ratio and moderate β phase volume fraction can contribute to higher superplastic elongations.

Experimental study of PHWR debris bed under boil-off condition

Abstract Experiments have been carried out to study the heatup behavior of a submerged debris bed constituted of segmented reactor channels of Pressurized Heavy Water Reactors (PHWR). The study has been done for under fully submerged, partially submerged and nearly exposed conditions. This situation may arise from a severe accident scenario of PHWR where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure tube and calandria tube constitute each segmented reactor channels. Ten such reactor channels constitute the debris bed. Heatup of this debris bed is observed with respect to different water levels and at power levels of 10, 20 and 30 kW, typical to decay power levels of 0.25 %, 0.5 % and 0.75 % of reactor full power respectively. It has been observed from the set of experiments that fully submerged debris bed does not get heated up and fuel rod temperature remains at saturation temperature. Heatup is observed for exposed channels only, however the heatup rate is limited with steam and radiative cooling. For exposed debris bed, channels surrounded by neighboring channels get heated up as compared to channels at the periphery of the debris bed.

Measurement of void fraction in flow boiling of ZnO–water nanofluids using image processing technique

In recent years, nanofluids have been an active area of research in many engineering applications, especially for nuclear reactor safety systems due to their enhanced thermal properties as a coolant. In this study, experiments were performed in subcooled flow boiling of water and ZnO–water nanofluids with different nanoparticle concentrations (0.001–0.01vol.%) in horizontal annulus at heat fluxes varying from 100 to 550kW/m2 and flow rates from 0.1 to 0.175lps at 1bar inlet pressure and constant subcooling of 20°C to determine the void fraction by image processing technique. Parametric effects of nanoparticle volume fraction, heat flux, flow rate and axial location of heater rod on void fraction were studied. Bubble images during flow boiling were captured with high speed visualization and analyzed by National Instruments IMAQ Vision Builder 6.1 image processing software. Results show that void fraction decreases up to 86% with the use of nanofluid in place of water and it also decreases with increasing nanoparticle concentration and flow rate, whereas increase in heat flux and axial location of heater rod have opposite effect.

Start-up transient test simulation with and without void-reactivity feedback for a two-phase natural circulation reactor

The new small-scaled light water reactor design, known as integral modular water reactor (IMR), has an advantage compared to conventional light water reactor in terms of costs by adopting innovative technologies. However, this reactor may be susceptible to flow instabilities due to two-phase natural circulation inside the reactor pressure vessel (RPV). Flow instabilities may be amplified due to strong interaction between the flow and core power through the void-reactivity feedback mechanism. During the start-up of the IMR, system pressure is low. At low pressure, the density ratio can be quite high, which leads to large variation in void fraction due to change in flow quality. In the IMR design, the long riser and large volume of water can lead to thermal non-equilibrium between the phases due to the significant variation of the saturation temperature along flow direction. This can result in flow oscillations at certain operating conditions.In order to understand and identify the instability phenomena during the start-up of the reactor, a scaled experimental facility is designed based on the sound scaling approach. The scaling laws are used to obtain design parameters to maintain the similarities between prototype and experimental facility. Modified power scaling is used for transient tests. Four heater rods are used to simulate the chaotic flashing phenomena. In order to investigate the instability during the start-up procedure, start-up transient tests are performed with and without void-reactivity feedback for different core heat-up rate. These experimental results show that IMR is susceptible to instability during the start-up procedure at very low pressure. These oscillations are condensation induced oscillations and occur due to high subcooling in chimney section. It is investigated that, since amplitude of power oscillation is low, void-reactivity feedback effect does not affect the stability of the system for this type of given power transient.