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In this study, upward air–water two-phase flow tests were carried out in a 3 cm diameter pipe under atmospheric pressure, and over 3000 data points were collected from a wide range of superficial gas and liquid velocities (⟨jg⟩ ≈ 0.02–30 m/s and ⟨jf⟩ ≈ 0.02–2 m/s) for the investigation of flow regime identification. The probability density function (PDF) of transient void fraction signals and its full-width at half-maximum (FWHM) were obtained and used for analysis and data classification. Considering the features of PDF profiles, the flow conditions can be classified into four regions, which show a fair agreement with the existing flow regime maps in general trends. Furthermore, by examining the FWHM distributions, two more regions with high-FWHM (HF) values were identified as the transitions of higher-flow bubbly-to-slug and slug-to-churn flows as well as most portion of churn flow, and a valley region next to the HF regions can express the transition of churn-to-annular flows. Overall, six groups of flow conditions can be classified based on the present methodology, and each group can be corresponding to specific flow regimes or transition regions. This study can provide a simple and efficient way for flow regime identification.

In this study, an upward air-water two-phase flow experiment was carried out with a transparent circular tube to investigate the flow characteristics of churn and annular flows. A simple and novel method utilizing the transient differential pressure and conductivity signals were proposed to identify the transition boundary of churn and annular flows. The test range covers the flow conditions of superficial gas and liquid velocities of Jg ≈ 5.1–30 m/s and Jf ≈ 0.04–0.4 m/s under the atmospheric pressure, and the statistical techniques including PDF, FFT and STD mesh plots were used to present the flow characteristics. According to the present experimental results, the churn flow may transform to annular flow at the transition boundary of Jg ≈ 12.6–15.7 m/s under the low liquid flow condition of Jf ⩽ 0.4 m/s. The present results were benchmarked with the existing flow regime maps including Mishima-Ishii and Taitel-Dukler maps, and the identified transition boundary of churn to annular flows shows a good agreement with the existing maps. The present method can provide a simple and efficient way to identify the flow regimes.

This study combines both concepts of solar ventilation technology and solar air collector. This is a quite innovative and potential facility to effectively use thermal energy and reduce the accumulation of heat in the indoor space simultaneously. The purpose of this study is to create a prototype and implement the experiments. Computational fluid dynamics (CFD) approach is employed to validate the characteristics of the flow and heat transfer. For the accuracy of numerical predictions, the method of Solar Ray Tracing was used for thermal radiation flux as boundary condition on the wall. The local heat transfer correlation was investigated to predict surrounding wind speed upon device cover. Three sorts of glasses and several aspect ratios of flow channels have been compared to conclude the optimal configuration. In addition, four important factors, such as the stagnant layer thickness, emissivity on the illustrated surface, mass flow rate and the height of the device, are also considered and discussed in detail. The result showed that the optimal design is dominated by the combination of an aspect ratio of 50 mm:10 mm, and appropriate mass flow rate to the height of the device. The present work on thermal energy collection can assist us in designing a powerful solar air collector in some potential applications.

In this study, a two-phase flow experiment has been carried out to investigate the feasibility of using probability density function for distinguishing dispersed bubbly flow from the flow regime map. The conductivity sensors were installed and used to acquire the transient conductivity signals, which have been converted to void fraction values during the tests. The Matlab software was used to draw the probability density function of void fraction, and it appears clear differences among the probability density function of void fraction in different flow patterns. Based on this, the probability density function is used to identify the boundary of dispersed bubble flow. By obtaining the important features of the probability density function profiles such as peak location, maximum/minimum values and area ratios of covered regions, the boundary of dispersed bubble flow can be identified.

According to “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants” (NUREG-0800) of the U.S. Nuclear Regulatory Commission, the homogeneous and thermal equilibriu...

A loss-of-coolant accident (LOCA) is one of the design basis accidents (DBAs) in nuclear power plant safety. DBAs may cause containment failure as a result of high temperature or pressure from LOCAs. This study investigates the integrity of the protective mechanism of the AP1000 containment system during a LOCA. The performance of the passive containment cooling system (PCCS) and the ability to perform decay heat removal for long-term cooling are evaluated in this study. The PCCS utilizes gravity-driven natural convection and naturally induced passive safety devices to move the decay heat of fuels to the atmosphere. In this study, two accidents are selected for analysis: a double-ended guillotine of a hot leg and a double-ended break of a main steam line. The analytical results are compared with the corresponding results provided in the AP1000 “Design Control Document” (DCD). The short-term calculations and comparisons with the DCD suggest that GOTHIC 7.2a with a natural convection model may properly represent the phenomenon of passive, safe heat removal. This paper also simulates the long-term calculations of spraying the outer primary containment with a water film and the steam condensation of the inner containment.

After the Fukushima Daiichi nuclear accident in 2011, the heat removal capability of the in-vessel retention external reactor vessel cooling (IVR-ERVC) process for the advanced light water reactors (ALWRs) becomes a popular topic of the nuclear safety analysis field. The IVR-ERVC is a strategy to mitigate the consequences of the severe nuclear accident if the reactor core melts and relocates to the lower head of the reactor pressure vessel (RPV). In order to understand the physical phenomenon at the outer cooling surface of the RPV, a new heat transfer experiment was designed and established to investigate the local phenomena of heat transfer, boiling, and critical heat flux (CHF) at the downward-facing surface. The test facility can simulate the partial heating by the high temperature debris of molten fuels, and the experiments were performed under pool and forced impinging flow conditions. Different injection distances from the coolant inlet to the heating surface (Pool boiling, 8, 12, 16, and 20 cm) were tested. Different inclining angles were also discussed in this study. With the test results, the shorter inlet distance performed the higher CHF limit. The larger inclining angle had performed higher CHF limit, because the larger inclining angle induced the bubbles departing faster and easier from the heating surface. A new CHF model with downward-facing inclined angles has been derived of this study.

The loss-of-coolant accident (LOCA) refers to the loss of a reactor coolant due to broken primary pipes in a nuclear power plant, which results in a rapid rise of containment temperature and pressure. The effused coolant flashes into steam because of the high core pressure, increasing the temperature and pressure inside the containment. To avoid the severe containment damage that may be caused by such a high temperature and pressure, a containment spray system is built at the top of the containment. When LOCA occurs, the core spray system will spray and condense the coolant to decrease the temperature and pressure inside the containment. However, as the operating time of the system increases, the water head of the pump built in the containment spray system decreases. A decrease in water head may result in reduction of water flow, and thus, suppress the capability of the spray system. Currently, there are limited studies on how head decline affects the internal pressure of a containment as well as on the safety measures to be employed when such a decrease in water head occurs. In this study, we are going to discuss whether containment spray systems are capable of adequately cooling and regulating the internal pressure of containments when experiencing a decrease in water head. In this study, an incident model LOCA caused by main streamline break was developed using GOTHIC. The heat loss of the wall was simulated by setting a heat sink. The model was validated by comparing it with the final safety analysis report (FSAR) of the Maanshan nuclear power plant (NPP).In this model, the containment spray system is composed of four height levels, each with different flow rates, based on the design of the Maanshan NPP. The boundary condition of the model was obtained by calculating the spraying flow rate of the containment spray system after the loss of water head using user-defined functions in computational fluid dynamics modeling. Then a precise model is built using GOTHIC. The simulation results show that the simulated flow rate is larger than that in the FSAR of Maanshan NPP. Even when the water head decreases, the peak pressure in the containment only increases by a small amount, which does not pose any potential damage to the containment.

The objective of this research is to develop a numerical model for the evaluation of the cooling capability of spent fuel pool (SFP) in removing decay heat. Computational fluid dynamics (CFD) approach has been applied to develop a three-dimensional two-phase thermal hydraulic model. To simplify the analysis, the pressure drop of the coolant flowing through the fuel region in the pool is modeled as a porous-medium. The geometry as well as the boundary conditions of SFP of second nuclear power plant in Taiwan has been employed as sample to be analyzed and simulated numerically. The results show that the current configuration has enough cooling capability to meet the licensing regulations under normal operation and most kinds of conditions. In the case of extremely inappropriate layout of spent fuel arrangement, the computed results illustrate that slight local saturation on fuel rods may occur unless external cooling system is supplied.

In this study the concept of generalized cogeneration has been employed to the analysis of the effective heat recovery from the high temperature gas-cooled reactor (HTGR). Brayton cycle is used as a topping cycle for HTGR with helium as the working fluid. HTR-10 and VHTR, having different power levels of HTGR, have been applied in the present study. The paper investigates the optimal strategy of the heat recovery, which combines the bottoming cycles and the parameters such as the select of different working fluids and the ambient temperatures. The heat recovery has integrated the organic Rankine cycle (ORC) using the fluids, such as alkane and benzene to perform the thermodynamic analysis. The total efficiency could be increased by 13.57% and 3.12% for HTR-10 and VHTR, respectively. Brayton cycle has been employed in the heat recovery for the iodine–sulfur (IS) hydrogen production process of VHTR by integrating the intermediate cooling process with the recuperator. When the sulfur trioxide (SO3) decomposition temperature is the lowest and the efficiency of the power conversion from hydrogen is assumed 0.6, the total thermal efficiency will be increased from 3.12% to 8.60%.

A thermo-hydraulic analysis model was developed to analyze thermal stratification phenomena observed in the hot-legs of pressurized water reactors (PWR). The model uses VIPREW code to determine the flow field and temperature distribution in the reactor fuel region. The temperature readings from the thermal couples located at the exit of the reactor core were used to compare with the VIPREW computed results. The predicted values agree well with the measurements. The VIPREW results are then used as the boundary conditions for the CFD analysis. The CFD computational domain includes the upper plenum and hot-legs and the fifty two (52) control rod guiding tubes to properly include the additional obstructions imposed to the fluid. Different fuel loading patterns were studied to investigate the effects of different power distribution and fuel channel exit water temperature on hot-leg thermal stratification magnitude. The analysis results show that the 52 control rod guide tubes have major contribution to the mixing effect in the upper plenum. The sudden expansion of the cross sectional area in the upper plenum leads to the formation of recirculation vortex that prolongs the duration of coolant in the reactor vessel. The hotter coolant from the center portion tends to flow upwards to the top before exiting at the upper portion of the hot-leg pipes. It leads to higher temperature in the upper portion of the hot-legs. Water from the cooler outer fuel channels tends to trap in the recirculation region before exiting from the lower portion of the hot-legs.

The two limiting pressurization transient cases for the Lungmen ABWR plant: (1) Load Rejection with No Bypass and (2) Single Control Valve Closure in Fast Closure Mode, have been analyzed using the RETRAN-02 analysis model, which is independent of the model used by the NSSS vendor. The calculation results have been compared with those shown in the Lungmen FSAR. In general, the comparison shows that the transient response of key system parameters agrees well with the FSAR results, including those relating to core power, core inlet flow, system pressure, core inlet enthalpy, steam flow rate, feedwater flow rate, reactor water level, etc. Based on this, it is expected that the RETRAN-02 model developed in this study can be employed to support the operational transient analysis if the model is benchmarked further with the appropriate data, e.g., the Lungmen plant power test data, etc.

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the ΔCPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

This study presents a numerical investigation of a whole set of thermal module, including a plate-fin heat sink embedded with a vapor chamber, subject to the influence of concentrated heat sources. Within the vapor chamber, the internal vapor is assumed as a common heat-transfer interface between the wicks. CFD simulations of the integrated heat sink are carried out with this assumption. The calculated results are in good agreement with the experiments, and show a maximum difference of 6.3% for the hotspot temperature rises. It is found that the area of the heat source has an important influence to the performance of the vapor chamber. The major spreading resistance of the vapor chamber comes from the bottom wall, where a concentrated heat source is applied. In addition, the isotropic and orthotropic approaches are proposed to calculate the effective thermal conductivities of the vapor chamber. By approximating the vapor chamber as a conduction plate, the effective conductivity can be obtained from the analytical solutions of the spreading resistances. The vapor chamber can reduce the spreading resistances sufficiently by its excellent lateral thermal spreading effect, which can be interpreted by the orthotropic approach.

This work numerically investigates the radiative flux and flame shape from heptane liquid pool fires with small to medium sized pool fires (14–38 cm), grid sizes, spectral bands and solid angles. Experimental results were originated for comparison. The results show that calculated flame shapes and radiative fluxes are in good agreement with the measurements for suitable grid resolution. Besides, the effect of the intermittency of a flame on thermal radiation was performed. Experimentally, a shield was used to shade the radiant power from a persistent flame, and the radiative flux was measured from an intermittent flame on 30 cm diameter pool fires. The results show that 36% of the radiative flux measured is emitted from the intermittent flame when the radiometer is located at the base of the pool fire. As the radiometer is moved upwards, the radiative fluxes measured from the intermittent flame increase gradually, even to 50%.

A steam generator (SG) plays a significant role not only with respect to the primary-to-secondary heat transfer but also as a fission product barrier to prevent the release of radionuclides. Tube plugging is an efficient way to avoid releasing radionuclides when SG tubes are severely degraded. However, this remedial action may cause the decrease of SG heat transfer capability, especially in transient or accident conditions. It is therefore crucial for the plant staff to understand the trend of plugged tubes for the SG operation and maintenance. Statistical methodologies are proposed in this paper to predict this trend. The accumulated numbers of SG plugged tubes versus the operation time are predicted using the Weibull and log–normal distributions, which correspond well with the plant measured data from a selected pressurized water reactor (PWR). With the help of these predictions, the accumulated number of SG plugged tubes can be reasonably extrapolated to the 40-year operation lifetime (or even longer than 40 years) of a PWR. This information can assist the plant policymakers to determine whether or when a SG must be replaced.

Flow accelerated corrosion (FAC) wear is a serious degradation problem especially for nuclear power plants since it may result in the plant damage as well as risk to personnel. In this paper, a methodology which includes the two-phase hydrodynamic CFD models and FAC models, is proposed to predict severe FAC wear sites. Based on hydrodynamic simulation results, the present CFD models can precisely capture the two-phase characteristics within the piping system, which include the centrifugal effect, the gravitation effect and the imbalance of phase and mass separation in a T-junction, etc. Coupled with these flow characteristics, the appropriate FAC indicators can predict the possible locations of severe FAC wear. This methodology was validated against the measured results of wear site distributions for the piping system in a boiling water reactor (BWR) power plant. Good agreement between measurements and predictions at severe wear sites indicate that the present models can capture the characteristics of severe FAC wear and can help assist in the pipe wall-monitoring program for a nuclear power plant.

In this paper, possible influences of power uprate on the distribution characteristics of erosion-corrosion (E/C) wear sites were analyzed through proper two-phase models. These models include thre...

In this study, the discrete ordinates method (DOM) model is employed to estimate the effect of thermal radiation from multiple heat sources in a natural-convection flow field. It is found that the flow field around the chips can be altered by natural convection as induced by radiative heat transfer. The influence of thermal radiation is higher than 65% when the chipboard is in a vertical orientation. Furthermore, even if the chip surface temperature is only 317 K, the influence of radiative heat transfer is still up to 18%. Therefore, radiative heat transfer cannot be ignored for electronic component computational fluid dynamics simulation under natural convection.

This study examines the spreading ability of rectangular plates numerically, analytically, and experimentally. The effect of aspect ratio, defined as an equivalent radius of a heater divided by that of a spreader plate, is investigated. The numerical results show a very good agreement with the analytical solutions. From the calculated results, the spreading resistance of the conduction plates with a small aspect ratio is higher than the one-dimensional conduction resistance. Calculated results also show that the spreading ability of a metal plate would be affected slightly by the external convective heat-transfer coefficient when the ratio of the longitudinal heat convection to the lateral heat spreading is less than 0.1. In addition to the numerical analysis, experimental comparisons between copper∕aluminum plates and a vapor chamber having the same thickness have been conducted. The experimental results show that the thermal resistance of the metal plates is independent of input power whereas that of the vapor chamber shows a noticeable drop with increased power. For the influence of concentrated heat source, the surface temperature distributions for the metal plates become concentrated with a reduced aspect ratio. However, the variations of the aspect ratio and the input power would yield minor effects to the surface temperature distribution of the vapor chamber. As compared with the conduction plates, the vapor chamber would offer a lower temperature rise and a more uniform temperature distribution. Thus, the vapor chamber provides a better choice as a heat spreader for concentrated heat sources.

The vapor chambers (flat plate heat pipes) have been applied on the electronic cooling recently. To satisfy the quick-response requirement of the industries, a simplified transient three-dimensional linear model has been developed and tested in this study. In the proposed model, the vapor is assumed as a single interface between the evaporator and condenser wicks, and this assumption enables the vapor chamber to be analyzed by being split into small control volumes. Comparing with the previous available results, the calculated transient responses have shown good agreements with the existing results. For further validation of the proposed model, a water-cooling experiment was conducted. In addition to the vapor chamber, the heating block is also taken into account in the simulation. It is found that the inclusion of the capacitance of heating block shows a better agreement with the measurements.

A new mechanistic critical heat flux (CHF) model has been developed for flow boiling CHF data of low-pressure (i.e., 0.2 to 4.0 MPa), low-mass-flux (i.e., 189 to 789 kg/m 2 .s), and high-quality conditions. In general, CHF at these conditions associates with the flow regime of annular flow. This model assumes that the Helmholtz instability at the liquid-vapor interface of annular flow triggers the onset of CHF. CHF is the energy required to dryout the liquid film isolated by flow instability With five empirical constants to properly correlate the liquid-vapor configurations of annular flow in the steam-water systems, the model successfully achieves a mean deviation error of 0.2% over a CHF data set consisting of 733 CHF data. The prediction of this model is more accurate than those of Biasi and Bowring correlations at the aforementioned low-pressure and low-mass-flux conditions.

During the past years, a number of reduced-scale test facilities have been constructed to investigate the physical phenomena of transients or accidents occurring in nuclear power plants. Since the behavior of a nuclear power plant is complicated, it is quite impossible for a small-scaled facility to simulate all the physical phenomena during the transient process. But, by way of proper scaling, most of the important aspects of transient behavior can be simulated. Calculations using RELAP5/MOD3 investigate whether most of the key thermal-hydraulic phenomena observed in the Institute of Nuclear Energy Research Integral System Test (IIST) facility can be expected in a prototype plant. When compared with experimental data, the calculated results of two different scale models show reasonable agreement with the natural circulation transients. The scale-up capability of RELAP5/MOD3 is demonstrated by simulating the single-phase and two-phase natural circulation transients. Also, the scaling distortions in the heat transfer areas of the IIST facility do not strongly distort the thermal-hydraulic behavior of experimental data.

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 analysis of r...

Wave propagation of low-void-fraction, two-phase bubbly flow is analyzed from the compressibility of a bubble and an assumption of homogeneity for steam bubbles. The phase velocity and attenuation calculated from the model are strongly dependent on frequency. The equilibrium sound velocity is derived for the limiting case as frequency approaches zero. The second resonance of a vapor bubble is also discussed. In addition, the void fraction, bubble size, and noncondensable gas effects are also analyzed in the model. In this paper the relationship between sound velocity and interfacial area density is derived, and a new method to measure interfacial area is proposed.

In this paper, wave propagation in a homogeneous, low void fraction, two-phase air-water bubbly flow is analyzed through the compressibility of a single bubble to derive a P({rho}) relation; the dispersion relation is then derived by a homogeneous model. The phase velocity and attenuation calculated from the model are compared with existing data and are in good agreement. The momentum transfer effect is considered through the virtual mass term and is significant at a higher void fraction. The interfacial heat transfer between phases is significant at low frequency, while bubble scattering effects are important at high frequency (near resonance). Bubble behavior at both low and high frequency is derived based on the isothermal and the adiabatic cases, respectively. The phase velocity occurs at the limiting condition in both cases. Furthermore, resonance is present in the model, and the resonant frequency is determined.

In this paper a new approach to bundle critical power predictions is presented. In addition to a very accurate critical heat flux (CHF) model, correction factors that account for the effects of grid spacers, heat flux non-uniformities, and cold walls, which are needed for critical power predictions for practical fuel bundles, are developed. By using the subchannel analysis code COBRA IIIC/MIT-1, local flow conditions needed as input to CHF correlations are obtained. Critical power is therefore obtained iteratively to ensure that the bundle power value from the subchannel analysis will cause CHF at only one point in the bundle. Good agreement with the experimental data is obtained. The accuracy is higher than that of the W-3 and EPRI-1 correlations for the limited data base used in this study. The effects of three types of fuel abnormalities, namely, local heat flux spikes, local flow blockages, and rod bowing, on bundle critical power are also analyzed. The local heat flux spikes and flow blockages have no significant influence on critical power. However, rod bowing phenomena have some effect, the severity of which depends on system pressure, the gap closure between adjacent rods, and the presence or absence of thimble tubes (cold walls).more » A correlation for the influence of various rod bowing phenomena on bundle critical power is developed. Good agreement with experimental data is shown.« less

A simple model is developed for predicting the dispersion and attenuation of sound in a homogeneous low void fraction air—water two-phase bubbly flow. The relationship between interfacial area density and sound attenuation is derived for dimensionless frequency, Ω (= ωR 0 2 /α l ) in the range 2000–8000. The interfacial area density is shown to be proportional to the square of the sound attenuation. Measurements of sound attenuation are performed for standing waves in a vertical waveguide containing an air—water bubbly mixture. The model of sound attenuation is compared with present and previous experimental data, and with previous theories. The interfacial area density predicted by the present model is compared with experimental data; the range of void fraction being 0.5–10% and the range of bubble radius being 0.6–2.2 mm.

Changes of flow-accelerated corrosion (FAC) wear sites on the piping due to the power uprate of nuclear power plant are investigated by way of computational fluid dynamics (CFD) models. These models proposed in this paper include the three-dimensional two-phase flow models and appropriate FAC models. The computations are performed using commercial code Fluent 6.2 which is control-volume-based. A boiling water reactor (BWR), located at Taiwan, is selected in the present analytical works. Simulation results clearly reveal that the present model can precisely capture the two-phase phenomena within the piping system. Coupled with the calculated two-phase flow characteristics, the appropriate FAC indictors can predict the local distributions of severe FAC sites. These predicted results show reasonable agreement with the plant measurements. Therefore, the impacts of power uprate on the changes of wear sites can be confidently investigated by the present CFD model. Through the comparisons of predictions for the selected BWRs under 100%, 105%, and 110% power levels, the simulation results clearly reveal that the power uprate does not significantly change the characteristics of FAC wear sites.Copyright © 2009 by ASME

Power uprate is currently being considered by several nuclear power plant (NPP) utilities as a way to improve plant economy since 1970s. In general there are three types of power uprate: measurement uncertainty recapture (

The development of practical and simple on-line measurement of two-phase mass flowrate is of prime interest to applied nuclear reactor safety research. Experiments were performed at a system pressure of 0.1 MPa. The highest value of the mass flowrate for air-water was 1233 kg m -2 s -1 , while the void fraction range was 0.1∼0.75, which was equivalent to 0.0004∼0.0023 for air quality X. For steam-water mixtures, where the mass flowrate was up to 1662 kg m -2 s -1 , the results showed good agreement, as did those for air-water two-phase flow. The measuring techniques are described briefly and a slip ratio associated with the drift-flux model is proposed to correlate the mass flowrate model.

This paper investigates methods using the signals detected by a single void fraction sensor to identify four kinds of typical vertical, cocurrent, upward, two-phase tube flow patterns. By analyzing 100 sets of time-varying void fraction signals acquired from an impedance device in an air-water two-phase loop, the results of the various methods are evaluated and demonstrated. With the high-frequency contribution fraction (HFCF) criteria, the success rate is 81%. An auxiliary criterion (the void fraction criterion) is proposed to increase the success rate to 92%. The results and the criteria from this study are compared with earlier studies. From the comparison, the applicability of the HFCF criterion to a system in which void fraction can be measured directly is verified.

The void fraction is one of the most important quantities in experimental studies of two-phase flow. Two simple and economical techniques to determine this quantity are developed and discussed. The...

In this study, a computational fluid dynamics model has been developed to explain and validate the experimental results originating from the concept of a substrate with an opening. It is found that the openings will interrupt the growth of the boundary layer on substrate surfaces and hence improve the cooling ability of a module without any additional active parts. Furthermore, the concept of openings has not only so far provided at least 12% improvement in heat transfer, but also reduced some difficulties in finding thermal solution, such as the manufacturing cost and the design freedom. More importantly, this study has provided a further step in the direction of demonstrating the opening effect.

A physically based theoretical model of gas-liquid and steam-water two-phase flow in a horizontal tee-junction is developed. The model includes five independent equations: the mixture continuity eq...

Due to the potential threat of reactor coolant system (RCS) overpressurization, loss-of-normal-feed-water (LONF) transients without reactor trip have received special attention in the analysis of pressurized water reactor (PWR) anticipated transients without trip (ATWT). The U.S. Nuclear Regulatory Commission requires every PWR to be equipped with an ATWT mitigation system actuation circuitry (AMSAC) so that the turbine will be tripped and auxiliary feedwater (AFW) added when an LONF transient occurs. An AMSAC design will be installed in both units of the Maanshan Nuclear Power Station (MNPS) to deal with ATWTs under LONF transient conditions. A best-estimate transient analysis performed with the RETRAN-02/MOD3 code is used to assess the safety function of the actuation circuitry designed for MNPS. Analytical results show that the peak RCS pressure will not exceed the 22.16-MPa safety limit if the moderator temperature coefficient is sufficiently negative and the actuation circuitry functions normally. Effects of the moderator temperature coefficient, the Doppler coefficient, pressurizer power-operated relief valves, effective time of the AFW system, the steam dump system, and the automatic control rod system are discussed.

A method based on noise analysis techniques that can be applied to the identification of two-phase flow patterns in nuclear reactors is proposed. The identifying criterion, the high-frequency contribution fraction (HFCF), offers new potential to the in-core recognition of two-phase flow patterns. By analyzing 76 sets of signals acquired from a research nuclear reactor where two-phase flow patterns are generated in an in-core air/water loop, the typical signal, autocorrelogram, and spectrum of each flow pattern are demonstrated and evaluated. The identification success rate is 87 or 93%, depending on whether churn flow is counted. A method to improve the identification rate is also presented. This study demonstrates that the fluctuation characteristics above 10 Hz are induced by two-phase flow itself and are independent of the driving source; thus, it is adequate to apply the HFCF to the identification of two-phase flow patters. This study shows that it is possible to identify two-phase flow patterns by HFCF values.

The thermal problems of most electronic components have been solved in recent years by installing various cooling equipments. The passive cooling, which has no need of the fluid-driving devices, can be more reliable and achieved with less cost. In this paper, the Taguchi’s method is applied on the optimization of the passive cooling for electronic systems, and a representative CFD (computational fluid dynamic) model is selectively implemented for statistical analysis. The selected parameters in this study are the power density, mother board orientation, chip geometry, opening between chips and the flow pattern. The optimal combination for thermal cooling is obtained by using a two-level statistical approach. Analysis results indicate that about 50% of the effort in performing experiments and simulations can be saved. Further, the results confirmed that the most important parameters affecting the thermal behaviors are the openings in the mother board, power density, and the flow pattern in sequence. Finally, the concept of opening increases the reliability, reduces the manufacturing cost, and simplifies the assembly procedures.

The performance of a modified design of local condensers to pre-treat a variety of volatile organic compounds (VOCs) emitted from the stripping process of a semiconductor fab was tested in this study. The reaction temperature of the condensers was controlled at around 10 °C, it is relatively higher than the traditional condenser reaction temperature. Both VOCs and water vapors were condensed and formed liquid films. This resulted in an enhancement of the VOCs removals, especially for VOCs of high boiling points or solubility. This can help to prevent the follow up zeolite concentrator from damage. The performance of the integrated system of condenser/zeolite concentrator could, therefore, remain highly efficient for a longer operation time. Its annualized cost would also be lower than installing the zeolite concentrator only.

The use of a honeycomb zeolite concentrator and an oxidation process is one of the most popular methods demonstrated to control volatile organic compound (VOCs) emissions from waste gases in semiconductor manufacturing plants. This study attempts to characterize the performance of a concentrator in terms of the removal efficiencies of semiconductor VOCs (isopropyl alcohol [IPA], acetone, propylene glycol methyl ether [PGME], and propylene glycol monomethyl ether acetate [PGMEA]) under several parameters that govern the actual operations. Experimental results indicated that at inlet temperatures of under 40 degrees C and a relative humidity of under 80%, the removal efficiency of a zeolite concentrator can be maintained well over 90%. The optimal rotation speed of the concentrator is between 3 and 4.5 rph in this study. The optimal rotation speed increases with the VOCs inlet concentration. Furthermore, reducing the concentration ratio helps to increase the removal efficiency, but it also increases the incineration cost. With reference to competitive adsorption, PGMEA and PGME are more easily adsorbed on a zeolite concentrator than are IPA and acetone because of their high boiling points and molecular weights.