The Experts below are selected from a list of 18186 Experts worldwide ranked by ideXlab platform
Jie Liu - One of the best experts on this subject based on the ideXlab platform.
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fuel and core design of super light water reactor with low leakage fuel loading pattern
Journal of Nuclear Science and Technology, 2006Co-Authors: Kazuhiro Kamei, Yuki Ishiwatari, Yoshiaki Oka, Akifumi Yamaji, Jie LiuAbstract:An equilibrium core for the High Temperature Supercritical-pressure Light Water Reactor, now called the Super Light Water Reactor (Super LWR), has been designed. The fuel assemblies loaded in the peripheral region of the core are cooled with descending Flow to achieve a high average coolant core outlet temperature. This Flow scheme is compatible with a low leakage fuel loading pattern (LLLP) in which 3rd cycle fuel assemblies are loaded in the core peripheral region. Stainless steel is used for fuel rod claddings and for structural materials. Watts correlations are used for predicting heat transfer in the core. They take into account the improved heat transfer for Downward Flow. It is found that the water rods with their Downward Flow need to be thermally insulated with thin ZrO2 layer to keep the moderator temperature below the pseudo critical temperature and to reduce the thermal stress in water rod walls. An average coolant core outlet temperature of 500°C is achieved. The effects of various heat trans...
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Safety of Super LWR, (II) : Safety Analysis at Supercritical Pressure
Journal of Nuclear Science and Technology, 2005Co-Authors: Yuki Ishiwatari, Yoshiaki Oka, Seiichi Koshizuka, Akifumi Yamaji, Jie LiuAbstract:This paper describes safety analysis of the high-temperature supercritical water-cooled thermal reactor with Downward-Flow water rods (called Super LWR) at supercritical pressure. Eleven transients and four accidents are chosen for the safety analysis considering types of abnormalities. The cladding temperature is taken as the important transient criterion instead of the heat flux ratio. The once-through cooling system and the Downward-Flow water rod system characterize safety of the Super LWR. “Loss of feedwater” is important because it is the same as “loss of reactor coolant Flow” unlike BWR and PWR. However, the Downward-Flow water rods mitigate core heat-up before startup of the auxiliary feedwater system because they remove heat from the fuel channels by heat conduction and supply their water inventory to the fuel channels by volume expansion. During pressurization transients, the reactor power does not increase significantly unlike BWR due to no void collapse in single-phase Flow and decrease in coo...
Oleg Zikanov - One of the best experts on this subject based on the ideXlab platform.
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Convection instability in a Downward Flow in a vertical duct with strong transverse magnetic field
Physics of Fluids, 2018Co-Authors: Xuan Zhang, Oleg ZikanovAbstract:The Downward Flow in a vertical duct with one heated and three thermally insulated walls is analyzed numerically using the two-dimensional approximation valid in the asymptotic limit of an imposed strong transverse magnetic field. The work is motivated by the design of liquid metal blankets with poloidal ducts for future nuclear fusion reactors, in which the main component of the very strong magnetic field is perpendicular to the Flow direction and very strong heating is applied at the wall facing the reaction chamber. The Flow is found to be steady-state or oscillating depending on the strengths of the heating and magnetic field. A parametric study of the instability leading to the oscillations is performed. It is found among other results that the Flow is unstable and develops high-amplitude temperature oscillations at the conditions typical for a fusion reactor blanket.The Downward Flow in a vertical duct with one heated and three thermally insulated walls is analyzed numerically using the two-dimensional approximation valid in the asymptotic limit of an imposed strong transverse magnetic field. The work is motivated by the design of liquid metal blankets with poloidal ducts for future nuclear fusion reactors, in which the main component of the very strong magnetic field is perpendicular to the Flow direction and very strong heating is applied at the wall facing the reaction chamber. The Flow is found to be steady-state or oscillating depending on the strengths of the heating and magnetic field. A parametric study of the instability leading to the oscillations is performed. It is found among other results that the Flow is unstable and develops high-amplitude temperature oscillations at the conditions typical for a fusion reactor blanket.
Bernardo Figueroa-espinoza - One of the best experts on this subject based on the ideXlab platform.
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Taylor bubble rising in a vertical pipe against laminar or turbulent Downward Flow: symmetric to asymmetric shape transition
Journal of Fluid Mechanics, 2014Co-Authors: Jean Fabre, Bernardo Figueroa-espinozaAbstract:The symmetry of Taylor bubbles moving in a vertical pipe is likely to break when the liquid flo ws Downward at a velocity greater than some critical value. The present experiments performed in the inertial regime for Reynolds numbers in the range 100 < Re < 10000 show that bifurcation to an eccentric motion occurs, with a noticeable increase of the bubble velocity. The influence of the surface tension parameter (an inverse Eotvos number), Σ, has been investigated for 0.0045 < Σ < 0.067. It appears that the motion of an asymmetric bubble is much more sensitive to surface tension than that of a symmetric bubble. For any given Σ, the symmetry-breaking bifurcation occurs in both laminar and turbulent Flow at the same vorticity-to-radius ratio (ω/r) 0 on the axis of the carrier fluid. This conclusion also applies to results obtained previously from numerical experiments in plane Flows. Taylor bubbles moving in a vertical pipe become asymmetric when the surrounding liquid Flows Downward at a mean velocity greater than some critical value. The tip of the long bullet-shaped bubbles moves into an eccentric position, and the resulting asymmetric bubbles move faster, relative to the liquid, than symmetric bubbles. This has a significant influence in slug Flow where most of the gas is carried in a series of long bubbles and this is why symmetry breaking was first observed in such Flow conditions. In their study of upward and Downward slug Flow Griffith & Wallis (1961) noted that ‘as the down Flow water velocity was increased, a point was reached at which the stable character of the bubble suddenly changes; the tip of the bubble began to distort to become alternately eccentric on one side or another’. Fifteen years later Martin (1976) focused on Downward slug Flow. His experimental results showed that ‘a stable T aylor bubble is only possible in Downward Flow for large values of the surface tension’ and that ‘the bubbles become more and more eccentric as the Downward Flow
Mamoru Ishii - One of the best experts on this subject based on the ideXlab platform.
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experimental study on vertical Downward air water two phase Flow in a large diameter pipe
International Journal of Heat and Mass Transfer, 2018Co-Authors: Guanyi Wang, M. Yousaf, Xiaohong Yang, Mamoru IshiiAbstract:Abstract Downward two-phase Flows in large diameter pipes are important in various industrial applications, especially for the safety analysis in nuclear reactors. To address the issue that few data of Downward Flow in large diameter pipes is available for model evaluation, experiments of air–water Downward Flow in a pipe with inner diameter of 203.2 mm have been performed. Area-averaged void fraction and pressure measurement, as well as Flow visualization, have been conducted at several axial locations. The Flow conditions for superficial gas velocity range from 0.05 m/s to 3.00 m/s and for superficial liquid velocity range from 0.1 m/s to 1.5 m/s, which cover cap-bubbly Flow, churn-turbulent Flow and annular/falling film Flow. The Flow structure at several axial locations and the transition from churn-turbulent Flow to annular/falling film Flow have been discussed. Current available drift-flux models developed for Downward Flow in regular pipes as well as for upward Flow in large pipes are evaluated using newly collected data. For churn-turbulent Flow, the data indicates a larger drift velocity than the model prediction. Corresponding drift-flux constitutive equations are suggested which can reduce the prediction error from 34.37% to 11.79%.
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Flow structure and Flow regime transitions of Downward two-phase Flow in large diameter pipes
International Journal of Heat and Mass Transfer, 2018Co-Authors: Guanyi Wang, M. Yousaf, Xiaohong Yang, Mamoru IshiiAbstract:Abstract Downward two-phase Flow in large diameter pipes appears in numerous industrial applications and nuclear reactor accidents. In this study, adiabatic air–water two-phase Flow experiments in a 203.2 mm diameter pipe have been conducted to investigate Flow regimes and their transitions in Downward and horizontal Flow. Three Flow regimes (cap-bubbly, churn-turbulent and annular Flow) were recognized in Downward Flow, as well three Flow regimes (stratified, plug and pseudo-slug Flow) were observed in horizontal section. Evolution of void fraction and Flow structure along the loop under different Flow conditions has been discussed. The Probability Density Function (PDF) and Cumulative Probability Density Function (CPDF) of area-averaged void fraction signals were utilized as the indicators for self-organized neural network (SONN) method to identify horizontal and vertical Downward Flow regimes, respectively. The Downward Flow regime maps for 203.2 mm diameter pipes have been proposed and compared with that for different diameter pipes. The results show that the Flow regime maps agree well with that of 101.6 mm, but don’t agree well with that of smaller diameter pipes (25.4 mm and 50.8 mm). It is found that the transition between churn-turbulent and annular Flow occurs at a certain superficial liquid velocity regardless of superficial gas velocity. A set of new transition criteria have been developed for Downward Flow regime transitions in large diameter pipes, and validated by the experimental data of 203.2 mm and 101.6 mm diameter pipes. Compared with existing models, these criteria provide more accurate predictions for Downward Flow regime transitions in large diameter pipes.
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Downward two phase Flow experiment and general Flow regime transition criteria for various pipe sizes
International Journal of Heat and Mass Transfer, 2018Co-Authors: Zijiang Yang, Mamoru Ishii, Xiaohong Yang, Zhuoran Dang, Jianqiang ShanAbstract:Abstract Flow regime map is often used in choosing constitutive correlations for the two-phase Flow model. The related research mainly concentrates on the vertical upward and horizontal Flow, while it is not sufficient in the vertical Downward Flow. Downward Flow is very important as it is frequently encountered in the industrial applications. To enrich the Downward Flow research, an experiment is performed on a piping system with an inner diameter of 0.1524 m. Four different Flow patterns (bubbly Flow, cap bubbly Flow, churn turbulent Flow, and annular Flow) are classified with the artificial neural network method. The probability density function (PDF) profile of each Flow pattern is discussed. The proposed Flow regime map is compared with the other experiments and the effect of the pipe size is discussed. The existing Downward Flow regime boundary criteria are assessed with the experiment results. It is found that these criteria cannot fit the experiment results well. A set of general boundary criteria are still needed. In this paper, the criteria for the boundary of the bubbly Flow, the boundary between the cap bubbly Flow and the slug Flow, and the boundary of the falling film regime are proposed. They are verified with the experiments on different size pipes. A significant inlet effect on the Flow regime boundary is found. The falling film boundary criterion proposed cannot be applied when a sparger is used to inject gas into the Downward test section.
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Local Liquid Velocity in Vertical Air-Water Downward Flow
Journal of Fluids Engineering, 2004Co-Authors: Xiaodong Sun, Mamoru Ishii, Sidharth Paranjape, Seungjin Kim, Hiroshi Goda, Joseph M. KellyAbstract:We present an experimental study of local liquid velocity measurement in Downward air-water bubbly and slug Flows in a 50.8 mm inner-diameter round pipe. The axial liquid velocity and its fluctuations were measured by a laser Doppler anemometry (LDA) system. The maximum liquid velocity in a Downward two-phase Flow could occur off the pipe centerline at relatively low liquid Flow rates and this observation is consistent with other researchers' results
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LDA measurement in air–water Downward Flow
Experimental Thermal and Fluid Science, 2004Co-Authors: Xiaodong Sun, Mamoru Ishii, Sidharth Paranjape, Jennifer UhleAbstract:Abstract Local characteristics of the liquid phase in air–water Downward Flow were investigated in a 50.8 mm inner-diameter round pipe. A laser Doppler anemometry (LDA) system was used to measure axial liquid velocity and its fluctuations. To reduce the measurement uncertainty, the experiments were performed in Flow conditions with low void fraction. Titanium dioxide particles with a mean diameter of 2 μm were used as seeding particles to enhance the data rate. Benchmark experiment in the single-phase liquid Flow was first carried out to ensure good performance of the LDA system in the current setup. A total of 13 Flow conditions were examined in air–water two-phase experiment. By applying a special setup of the LDA system, it was found that no further signal discrimination process was required to obtain the liquid velocity in the present low void fraction conditions. The comparisons between the liquid Flow rates measured by the magnetic Flow meter and those obtained from the local measurements showed good agreements, with differences less than 6.0%. The measurement results demonstrated that the presence of the bubbles tended to flatten the liquid velocity radial profile, and the maximum liquid velocity might occur off the pipe centerline, in particular at relatively low Flow rates. Furthermore, the axial liquid velocity fluctuations were quite uniform in the radial direction. No significant turbulent reduction in the two-phase Downward Flow was observed in the current experimental Flow conditions.
Yuki Ishiwatari - One of the best experts on this subject based on the ideXlab platform.
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thermal and stability considerations for a supercritical water cooled fast reactor with Downward Flow channels during power raising phase of plant startup
Nuclear Engineering and Design, 2009Co-Authors: Yuki Ishiwatari, Satoshi IkejiriAbstract:Abstract This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant Flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by Downward Flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The Flow rate distribution as the function of both the reactor thermal power and the feedwater Flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the Flow rate distribution, thermal analyses and thermal–hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater Flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal–hydraulic stability in three “hot” channels; two seed assemblies with upward/Downward Flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.
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fuel and core design of super light water reactor with low leakage fuel loading pattern
Journal of Nuclear Science and Technology, 2006Co-Authors: Kazuhiro Kamei, Yuki Ishiwatari, Yoshiaki Oka, Akifumi Yamaji, Jie LiuAbstract:An equilibrium core for the High Temperature Supercritical-pressure Light Water Reactor, now called the Super Light Water Reactor (Super LWR), has been designed. The fuel assemblies loaded in the peripheral region of the core are cooled with descending Flow to achieve a high average coolant core outlet temperature. This Flow scheme is compatible with a low leakage fuel loading pattern (LLLP) in which 3rd cycle fuel assemblies are loaded in the core peripheral region. Stainless steel is used for fuel rod claddings and for structural materials. Watts correlations are used for predicting heat transfer in the core. They take into account the improved heat transfer for Downward Flow. It is found that the water rods with their Downward Flow need to be thermally insulated with thin ZrO2 layer to keep the moderator temperature below the pseudo critical temperature and to reduce the thermal stress in water rod walls. An average coolant core outlet temperature of 500°C is achieved. The effects of various heat trans...
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Safety of Super LWR, (II) : Safety Analysis at Supercritical Pressure
Journal of Nuclear Science and Technology, 2005Co-Authors: Yuki Ishiwatari, Yoshiaki Oka, Seiichi Koshizuka, Akifumi Yamaji, Jie LiuAbstract:This paper describes safety analysis of the high-temperature supercritical water-cooled thermal reactor with Downward-Flow water rods (called Super LWR) at supercritical pressure. Eleven transients and four accidents are chosen for the safety analysis considering types of abnormalities. The cladding temperature is taken as the important transient criterion instead of the heat flux ratio. The once-through cooling system and the Downward-Flow water rod system characterize safety of the Super LWR. “Loss of feedwater” is important because it is the same as “loss of reactor coolant Flow” unlike BWR and PWR. However, the Downward-Flow water rods mitigate core heat-up before startup of the auxiliary feedwater system because they remove heat from the fuel channels by heat conduction and supply their water inventory to the fuel channels by volume expansion. During pressurization transients, the reactor power does not increase significantly unlike BWR due to no void collapse in single-phase Flow and decrease in coo...