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[en] Highlights: • Numerical study on the optimization of windbreak structure for IDAC was conducted. • Windbreak wall is the most effective structure but is affected by wind direction. • The louver is next best and it can be flexibly adjusted at the windy conditions. • An optimal louver opening was obtained for achieving a good cooling performance. - Abstract: The heat transfer performance of indirect air-cooling (IDAC) towers in large power stations is sensitive to the ambient wind velocity. To ensure the economic and reliable operation of units under windy conditions, it is important to conduct research on the optimization of different wind-break structures. This paper uses computational fluid dynamics method (CFD) to simulate the heat transfer performance of a 1000 MW IDAC tower power stations with four different wind-break structures namely, cross walls, wind-break walls, cross line-screen, and louvers. The research results show that the order of the effective heat transfer improvement of four wind-break structures is the wind-break, cross wall, line-screen and louvers. The wind-break wall is the most optimal structure, but its performance is strictly influenced by the direction and velocity of the wind, and the cross walls and cross line-screen structure have similar limitation in the practice operation. The louver is installed in each sector, and it is the next best option for increasing the heat transfer performance. It can be flexibly adjusted based on the wind direction and velocity. With the decrease in the louver opening, k, there is a decrease in the heat transfer rate of the windward sectors, and a significant increase in the heat transfer rate of the leeward sectors. Thus the total heat transfer rate of the IDAC tower can be improved tremendously. Based on the analysis of heat transfer and air flow mechanisms, there is an optimal opening, k, which achieves the largest heat transfer performance in an IDAC tower at each wind velocity. This study provides an effective and practical approach to improve the efficiency of power plants in a windy area.
[en] Highlights: • The law of dynamic distribution of the heat and flow load are studied using three experiments. • The strain response of the pressure vessel to temperature and pressure load is obtained. • The frequency range of strong impact on structural response is obtained. - Abstract: An emergency core cooling system for a pressurized water reactor adopts direct safety injection with reactor pressure vessel. In this design, a special flow guide device is introduced in order to minimize the heat effects on the reactor internals. But this design makes the pressure vessel bear stronger heat and current impact. To investigate the dynamic response of the pressure vessel during safety injection, the law of dynamic distribution of the heat and flow load are studied using three experiments: In the visualization experiment, the relationship between injection condition and distribution pattern in the downcomer is obtained. In the heat mixing experiment, measuring the temperature and pressure near the inner wall of the pressure vessel, enables us to find out the law that governs dynamic distribution of the heat and pressure load as well as the main distribution area of these loads, and analyze how the temperature oscillation generated. In the structural response experiment, the strain response of the pressure vessel to temperature and pressure is obtained. Moreover, the frequency range of its response to hot oscillation under safety injection is also obtained by analysis. This study provides support to recognize the action law of heat, pressure and structural response in the reactor during safety injection.
[en] In order to analyze and evaluate the effect of pipe type on thermal stratification for PWR pressurizer surge line, computational fluid dynamics (CFD) method was used for the current and improved screw surge line during heating up condition. The maximum temperature differences and velocity distribution along the surge line axis under different velocities were obtained. Comparative analysis indicates that replacing current smooth tube with screw pipe can enhance the turbulent flow and the mixing effects between the hot and cold fluids. The temperature differences in flow cross sections along the screw surge line axis are reduced by about 1/3 compared with current smooth surge line, yielding effectively mitigation to the influence of thermal stratification. (authors)
[en] Experiments on heat transfer and characteristic of water in circumferential non-uniformly heated, vertical up flow tube at near critical pressure have been carried out at Xi'an Jiaotong University. The wall temperature and heat flux distribution along the cross section of the tube were obtained in the experiments. The critical heat flux, critical mass flow rate and critical mass quality have been analyzed. The effect of the heating method on the heat transfer characteristic has also been obtained. Experimental conditions included pressures of 19-22 MPa, mass flow velocities of 600-1200 kg/m2 s, and heat fluxes of 200-600 kW/m2. The experimental results showed that, for the non-uniformly heated tube, film boiling and nucleate boiling can coexist simultaneously at the different sides of one cross section of the tube when heat transfer deterioration occurs at high flux side. When the heat transfer deterioration occurs, the local critical peak heat flux of the non-uniformly heated tube and the critical heat flux of the uniformly heated tube are almost the same. So, the heat transfer deterioration is caused by the local heating condition of the non-uniformly heated tube. The minimum heat transfer coefficient of the non-uniformly and of the uniformly heated vertical tubes differs only a little when heat transfer deterioration occurs. Finally, the experimental correlations used to calculate the critical heat flux and the minimum heat transfer coefficient were given. (authors)
[en] Highlights: • A model was developed to analyze spray flash evaporation. • The temperature variations against distance are obtained from the model. • The model shows good agreement with experimental results. • The effects of influencing factors are studied. - Abstract: To enable a more in-depth understanding of the flash evaporation from the downward jet and extend the research range of spray flash evaporation, a mathematical model based on the diffusion-controlled evaporation model was proposed and developed. The droplet motion, droplet size variation and temperature variation are taken into account in the present model. The model was validated against the experimental data sets from literature sources. The temperature variation against the traveled distance was obtained and analyzed. Four variables, namely, flow velocity, pressure attenuation ratio, droplet size and relative humidity, were investigated by means of this model.
[en] Highlights: • Thermal stratification in the tilted surge line was experimentally studied. • Thermal stratification with different arrangements was compared. • Both the outer and the inner temperature distributions were analyzed. - Abstract: Thermal stratification in a pipe, which can lead to temperature fluctuations, has attracted much attention from researchers who study nuclear power plants. In particular, temperature fluctuation in the pressurizer surge line has been assessed as being a significant technical and safety issue due to its high rate of occurrence. The present work investigated thermal stratification of the tilted arrangement. Both the outer and the inner temperature distribution were extracted and analyzed. As the surge line inlet velocity increased, the regions where thermal stratification occurred gradually approached the hot leg. In addition, thermal stratification for the tilted arrangement was compared with that for the horizontal arrangement. The temperature difference along the surge line and the temperature patterns for the different arrangements were compared and analyzed.
[en] Highlights: • The function of a special direct vessel injection structures is tested. • Experiment and numerical study for flow mixing and heat transfer in reactor vessel are performed. • Capability of different turbulent numerical models are assessed. • A certain critical region is identified on reactor vessel surface. • Temperature and heat transfer coefficient distribution are obtained at reactor vessel. - Abstract: A 1400 MW pressurized-water reactor has adopted a special direct vessel injection (DVI) structure for the emergency core cooling system. This design makes the flow mixing and heat transfer in the reactor vessel very complicated and is very different from the traditional structure. This study focused on flow mixing and heat transfer capability in the reactor vessel under different injection conditions. The computational fluid dynamics method and two reference experiments were used. The research presents a numerical way that can provide sufficient accuracy. The DVI deflector was proved to protect the reactor internals from direct scouring by cold injection liquid, as designed. A key area was identified on the reactor vessel surface, where the heat transfer quantity and temperature gradient were obviously higher. And the dimensionless temperature and heat transfer capability were obtained at the reactor vessel. Such data can be used in future real scale reactor design. The study complements the research in understanding the origins of thermal fatigue and pressurized thermal shock in the reactor vessel, and the ability to quantify them.
[en] In ultra-supercritical pressure region, the heat transfer characteristics of water in a vertical upward internally ribbed tube with the diameter of Φ28.6 x 5.8 mm have been experimentally investigated. The tests have been performed under various conditions with the pressure ranging from 25 to 35 Mpa, the mass velocity from 450 to 1800 kg/(m2·s), and the internal wall heat flux from 200 to 600 kW/m2. The results show that the heat transfer of ultra-supercritical water is better below the pseudo-critical point than that above the pseudo-critical point. Increasing the mass velocity can improve the heat transfer of ultra-supercritical water with a much stronger effect below the pseudo-critical point than that above the pseudo-critical point. The pressure has only a moderate effect on the heat transfer of ultra-supercritical water when the temperature of water is below the pseudo-critical point. Sharp rising of inner wall temperature near the pseudo-critical region takes place earlier at a higher pressure. For given pressure and mass velocity, the internal wall heat flux also shows a significant effect on the distributing of inner wall temperature. The inner wall temperatures are higher at a higher heat flux. Increasing of internal wall heat flux leads to an early occurrence of sharp rising of the wall temperature. Based on the experiments, correlations of heat transfer coefficients are also presented for the vertical upward internal ribbed tubes. (authors)
[en] A supercritical water heat transfer test section has been built at Xi'an Jiaotong University to study the heat transfer from a 10 mm rod inside a square vertical channel with a wire-wrapped helically around it as a spacer. The test section is 1.5 m long and the wire pitch 200 mm. Experimental conditions included pressures of 23-25 MPa, mass fluxes of 500-1200 kg/m2 s, heat fluxes of 200-800 kW/m2, and inlet temperatures of 300-400 oC. Wall temperatures were measured with thermocouples at various positions near the rod surface. The experimental Nusselt numbers were compared with those calculated by empirical correlations for smooth tubes. The Jackson correlation showed better agreement with the test data compared with the Dittus-Boelter correlation but overpredicted the Nusselt numbers almost within the whole range of experimental conditions. Both correlations cannot predict the heat transfer accurately when deterioration occurred at low mass flux and relatively high heat flux in the pseudocritical region. Comparison of experimental data at two different supercritical pressures showed that the heat transfer was more enhanced at the lower supercritical pressure but the deterioration was more likely to occur at the higher pressure, meaning increased safety. Based on a comparison with an identical channel without the helical wrapped wire, it was found that the wire spacer does not enhance the heat transfer significantly under normal heat transfer conditions, but it contributes to the improvement of the heat transfer in the pseudocritical region and to a downstream shift of the onset of the deterioration. The Jackson buoyancy criterion is found to be valid and works well in predicting the onset of heat transfer deterioration occurring in the experiments without wire.
[en] Highlights: • The CFD results under various turbulent models were compared to experiment results. • The SST turbulent model was adopted for simulation of thermal stratification. • The temperature difference of the cross-section along the surge line was studied. - Abstract: Thermal stratification is a common phenomenon in the surge line of Pressurized Water Reactors (PWRs). A study using Computational Fluid Dynamics (CFD) has been performed to analyze the thermal stratification in the pressurized surge line with different surge line and main pipe velocity conditions. This study is typical of considering both the flow subjected to the buoyancy and the conjugate heat transfer between the fluid and the pipe wall. What makes the research complex is exactly the impact of buoyancy on thermal stratification. To validate the accuracy of CFD calculations, necessary comparison is performed with experiment results. Models with different computational conditions resulting from various similarity criterion parameters are used. More meaningful information than that from the experimental results can be obtained. At the same time, many actual operating conditions which are difficult for experiment can be also accomplished by applying CFD mode. Temperature distributions inside and on the outer surface of the surge line are provided by the CFD mode with different turbulent models. Compared to the standard k-ɛ turbulence model and Reynolds stress model (RSM), the SST k-ω turbulent model is more adapted for this study. Based on the conclusion, the numerical results with the SST k-ω turbulence model for the prototype on the actual nuclear power plant unit were mainly described.