Present study is focused on the computational analysis of melting of PCM inside the spherical capsule. Both unconstrained and constrained melting is analyzed for the constant PCM volume and similar initial and boundary conditions. RT27 is chosen as the PCM for this study. Air is considered at the top of PCM inside the spherical capsule. Results are validated with the existing experimental and computational results and found to be in good agreement. Results obtained from present study are compared for the melting fraction, pattern and time. Composite diagrams are presented for the streamline and temperature contours.
{"title":"A Comparative Study of Constrained and Unconstrained Melting Inside a Sphere","authors":"R. Kothari, S. Revankar, S. Sahu, S. I. Kundalwal","doi":"10.1115/icone2020-16056","DOIUrl":"https://doi.org/10.1115/icone2020-16056","url":null,"abstract":"\u0000 Present study is focused on the computational analysis of melting of PCM inside the spherical capsule. Both unconstrained and constrained melting is analyzed for the constant PCM volume and similar initial and boundary conditions. RT27 is chosen as the PCM for this study. Air is considered at the top of PCM inside the spherical capsule. Results are validated with the existing experimental and computational results and found to be in good agreement. Results obtained from present study are compared for the melting fraction, pattern and time. Composite diagrams are presented for the streamline and temperature contours.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"232 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132702845","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Chen Jiarui, Liu Jianchang, Li Dongyang, Tan Si-chao
As the key equipment to control the pressure stability of the coolant system, the pressurizer plays a role in maintaining the primary system pressure in the reactor. During the operation of the sea-based reactor, the internal free liquid level of the pressurizer will fluctuate greatly with different marine cycles, causing additional acceleration in the horizontal or vertical direction, which will cause the water level measured by the differential pressure measurement method to deviate from the actual water level. It will adversely affect the judgment and control of the signal. Moreover, the fluctuating liquid level will frequently trigger the water level alarm signal, resulting in the submersion of the sprinkler tuber and the exposure of the electric heating rod, which will reduce the safety and economy of the reactor. Therefore, this research is aimed at suppressing the fluctuation range of the water level and correcting the deviation of the water level measurement so as to improve the inherent safety of the reactor. In the present study, the experimental system consists of a motion excitation drive mechanism and an optical system. The experimental system has successfully established sloshing phenomenon of the pressurizer under different forms of motion by Laser induced fluorescence (LIF) technique and the experimental results obtained are compared with numerical results. The results of the research show that the pressurizer can make significant free surface fluctuation when excitation close to the natural frequency of the pressurizer. The suppression model developed by FLUTENT can effectively reduce the fluctuation range of free liquid level. In addition, the deviation of water level measurement enlarges with the swing angle increasing. The deviation can be reduced to the allowable error range by means of angle correction.
{"title":"Study on Effect of Sloshing Phenomenon on Water Level of Pressurizer","authors":"Chen Jiarui, Liu Jianchang, Li Dongyang, Tan Si-chao","doi":"10.1115/icone2020-16967","DOIUrl":"https://doi.org/10.1115/icone2020-16967","url":null,"abstract":"\u0000 As the key equipment to control the pressure stability of the coolant system, the pressurizer plays a role in maintaining the primary system pressure in the reactor. During the operation of the sea-based reactor, the internal free liquid level of the pressurizer will fluctuate greatly with different marine cycles, causing additional acceleration in the horizontal or vertical direction, which will cause the water level measured by the differential pressure measurement method to deviate from the actual water level. It will adversely affect the judgment and control of the signal. Moreover, the fluctuating liquid level will frequently trigger the water level alarm signal, resulting in the submersion of the sprinkler tuber and the exposure of the electric heating rod, which will reduce the safety and economy of the reactor. Therefore, this research is aimed at suppressing the fluctuation range of the water level and correcting the deviation of the water level measurement so as to improve the inherent safety of the reactor. In the present study, the experimental system consists of a motion excitation drive mechanism and an optical system. The experimental system has successfully established sloshing phenomenon of the pressurizer under different forms of motion by Laser induced fluorescence (LIF) technique and the experimental results obtained are compared with numerical results. The results of the research show that the pressurizer can make significant free surface fluctuation when excitation close to the natural frequency of the pressurizer. The suppression model developed by FLUTENT can effectively reduce the fluctuation range of free liquid level. In addition, the deviation of water level measurement enlarges with the swing angle increasing. The deviation can be reduced to the allowable error range by means of angle correction.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125209545","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Tomomasa Ito, Yuta Watanabe, Yasunori Yamamoto, Nassim Sahboun, Shuichiro Miwa
Understanding the spreading behavior of the molten core is important for predicting the progress of severe accidents and for smooth decommissioning work. In this study, we performed molten metal drop experiments using Sn and Cu to obtain fundamental data and its uncertainties that can be used to verify the analysis model and to evaluate the effect of scaling parameters on the spreading and deposition behaviors. We obtained spreading area and deposition thickness of the metals after floor impingement. The standard deviation on average were 6.0% and 6.7% for the thickness and the area, respectively. In addition, the correlation with each dimensionless number was confirmed, and strong correlations were obtained between the deposition behavior and the Re number, and between the spreading behavior and the Pe number. We constructed a correlation equation using our experimental data as a function of Re number, We number, Fr number, Pe number, and dimensionless length (L/d). The errors for our experimental data were within 30%.
{"title":"Experimental Study for Evaluation of Spreading Behavior of Free-Falling Molten Core With Floor Impingement","authors":"Tomomasa Ito, Yuta Watanabe, Yasunori Yamamoto, Nassim Sahboun, Shuichiro Miwa","doi":"10.1115/icone2020-16942","DOIUrl":"https://doi.org/10.1115/icone2020-16942","url":null,"abstract":"\u0000 Understanding the spreading behavior of the molten core is important for predicting the progress of severe accidents and for smooth decommissioning work. In this study, we performed molten metal drop experiments using Sn and Cu to obtain fundamental data and its uncertainties that can be used to verify the analysis model and to evaluate the effect of scaling parameters on the spreading and deposition behaviors. We obtained spreading area and deposition thickness of the metals after floor impingement. The standard deviation on average were 6.0% and 6.7% for the thickness and the area, respectively. In addition, the correlation with each dimensionless number was confirmed, and strong correlations were obtained between the deposition behavior and the Re number, and between the spreading behavior and the Pe number. We constructed a correlation equation using our experimental data as a function of Re number, We number, Fr number, Pe number, and dimensionless length (L/d). The errors for our experimental data were within 30%.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127503514","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yujia Liu, Sifan Peng, N. Gui, Xingtuan Yang, J. Tu, Shengyao Jiang
The pebbles flow is a fundamental issue for both academic investigation and engineering application in reactor core design and safety analysis. In general, experimental methods including spiral X-ray tomography and refractive index matched scanning technique (RIMS) are applied to obtain the identification of particles’ positions within a three-dimensional pebble bed. However, none of the above methods can perform global bed particles’ position identification in a dynamically discharging pebble bed, and the corresponding experimental equipment is difficult to access due to the complication and high expense. In this research, the experimental study is conducted to observe the gravity driven discharging process in the quasi two-dimensional silos by making use of the high-speed camera and the uniform backlight. A mathematical morphology-based method is applied to the pre-processing of the captured results. After being increased the gray value gradient by the threshold segmentation, the edges of the particles are identified and smoothed by the Sobel algorithm and the morphological opening operation. The particle centroid coordinates are identified according to the Hough circle transformation of the edges. For the whole pebble bed, the self-programmed process has a particle recognition accuracy of more than 99% and a particle centroid position deviation of less than 3%, which can accurately obtain the physical positions of all particles in the entire dynamically discharge process. By analyzing the position evolution of individual particles in consecutive images, velocity field and motion events of particles are observed. The discharging profiles of 5 conditions with different exit are analyzed in this experiment. The results make a contribution to improving the understanding of the mechanism of pebbles flow in nuclear engineering.
{"title":"Experimental Study on Gravity Driven Discharging of Quasi-Two-Dimensional Pebble Bed Based on Mathematical Morphology","authors":"Yujia Liu, Sifan Peng, N. Gui, Xingtuan Yang, J. Tu, Shengyao Jiang","doi":"10.1115/icone2020-16367","DOIUrl":"https://doi.org/10.1115/icone2020-16367","url":null,"abstract":"\u0000 The pebbles flow is a fundamental issue for both academic investigation and engineering application in reactor core design and safety analysis. In general, experimental methods including spiral X-ray tomography and refractive index matched scanning technique (RIMS) are applied to obtain the identification of particles’ positions within a three-dimensional pebble bed. However, none of the above methods can perform global bed particles’ position identification in a dynamically discharging pebble bed, and the corresponding experimental equipment is difficult to access due to the complication and high expense.\u0000 In this research, the experimental study is conducted to observe the gravity driven discharging process in the quasi two-dimensional silos by making use of the high-speed camera and the uniform backlight. A mathematical morphology-based method is applied to the pre-processing of the captured results. After being increased the gray value gradient by the threshold segmentation, the edges of the particles are identified and smoothed by the Sobel algorithm and the morphological opening operation. The particle centroid coordinates are identified according to the Hough circle transformation of the edges. For the whole pebble bed, the self-programmed process has a particle recognition accuracy of more than 99% and a particle centroid position deviation of less than 3%, which can accurately obtain the physical positions of all particles in the entire dynamically discharge process. By analyzing the position evolution of individual particles in consecutive images, velocity field and motion events of particles are observed. The discharging profiles of 5 conditions with different exit are analyzed in this experiment. The results make a contribution to improving the understanding of the mechanism of pebbles flow in nuclear engineering.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"98 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121442819","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yasunori Yamamoto, Masayoshi Mori, K. Ono, Tetsuya Takada
Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.
{"title":"Experimental Study of the Effect of Hydrogen Inflow on Passive Core Cooling System With Natural Circulation Flow","authors":"Yasunori Yamamoto, Masayoshi Mori, K. Ono, Tetsuya Takada","doi":"10.1115/icone2020-16949","DOIUrl":"https://doi.org/10.1115/icone2020-16949","url":null,"abstract":"\u0000 Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127760893","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Lead and Bismuth Eutectic (LBE) cooled fast reactor is attracting attention due to its advantages in safety. A thermal hydraulic analysis model is developed for an LBE cooled fast reactor helical coiled type steam generator, which could be used for predicting thermal hydraulic parameters distribution along the length of the tube. Based on two fluid model, the mathematical heat transfer model includes six regions: single water and vapor, subcooled boiling, saturated boiling, transition boiling and film boiling. In order to describe the heat and mass transfer between phases, the interphase model is presented. The steam generator is simplified with single tube concept and all the flow variables are evaluated at each point of the grid in which the domain is discretized. Full load condition and a postulated scenario, such a flow rate step changed in secondary side, is calculated in this study. The results showed that the changing process of the thermal hydraulic parameters of helically coiled steam generator conforms to the qualitative mechanism analysis results of thermal hydraulic analysis.
{"title":"Modeling of Thermal Hydraulic Characteristics for a LBE-Cooled Fast Reactor Helical Coiled Type Steam Generator","authors":"Xue Ding, Q. Wen, Zhiqiang Chen, Shenhui Ruan, Cheng Cheng","doi":"10.1115/icone2020-16741","DOIUrl":"https://doi.org/10.1115/icone2020-16741","url":null,"abstract":"\u0000 Lead and Bismuth Eutectic (LBE) cooled fast reactor is attracting attention due to its advantages in safety. A thermal hydraulic analysis model is developed for an LBE cooled fast reactor helical coiled type steam generator, which could be used for predicting thermal hydraulic parameters distribution along the length of the tube. Based on two fluid model, the mathematical heat transfer model includes six regions: single water and vapor, subcooled boiling, saturated boiling, transition boiling and film boiling. In order to describe the heat and mass transfer between phases, the interphase model is presented. The steam generator is simplified with single tube concept and all the flow variables are evaluated at each point of the grid in which the domain is discretized. Full load condition and a postulated scenario, such a flow rate step changed in secondary side, is calculated in this study. The results showed that the changing process of the thermal hydraulic parameters of helically coiled steam generator conforms to the qualitative mechanism analysis results of thermal hydraulic analysis.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"78 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126297626","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Z. Lei, Jian Deng, Wei Li, Xiaoli Wu, Deng Chunrui
Core melting and molten migration behavior are hot and difficult issues in the field of nuclear reactor severe accident research. The Moving Particle Semi-implicit (MPS) meshless method has potential to simulate free-surface and multiphase flows. In this study, the MPS method was utilized to simulate the melting process of UO2-Zr rod-type fuel elements. The models of heat conduction with phase change, simplified UO2-Zr eutectic reaction, viscous flow and surface tension were implemented with the framework of standard MPS method. Then, the improved MPS code was used to simulate and analyze the process of high-temperature melting and characteristics of molten migration and solidification in the coolant channel, aiming at revealing the severe accidents for light water reactors (LWR), particularly the early core damage. The results showed that compared with the case of higher initial temperature, when the initial temperature of molten UO2 is lower, more molten UO2 will solidify on the surface of rod cluster, and the blockage of upper flow channel caused by molten UO2 is more serious. In addition, this study also demonstrated the potential of the MPS method for the study of complicated severe accident phenomena in not only traditional LWR but also advanced nuclear reactors in the future.
{"title":"Numerical Study of Fuel Melting and Molten Migration Based on the MPS Method","authors":"Z. Lei, Jian Deng, Wei Li, Xiaoli Wu, Deng Chunrui","doi":"10.1115/icone2020-16806","DOIUrl":"https://doi.org/10.1115/icone2020-16806","url":null,"abstract":"\u0000 Core melting and molten migration behavior are hot and difficult issues in the field of nuclear reactor severe accident research. The Moving Particle Semi-implicit (MPS) meshless method has potential to simulate free-surface and multiphase flows. In this study, the MPS method was utilized to simulate the melting process of UO2-Zr rod-type fuel elements. The models of heat conduction with phase change, simplified UO2-Zr eutectic reaction, viscous flow and surface tension were implemented with the framework of standard MPS method. Then, the improved MPS code was used to simulate and analyze the process of high-temperature melting and characteristics of molten migration and solidification in the coolant channel, aiming at revealing the severe accidents for light water reactors (LWR), particularly the early core damage. The results showed that compared with the case of higher initial temperature, when the initial temperature of molten UO2 is lower, more molten UO2 will solidify on the surface of rod cluster, and the blockage of upper flow channel caused by molten UO2 is more serious. In addition, this study also demonstrated the potential of the MPS method for the study of complicated severe accident phenomena in not only traditional LWR but also advanced nuclear reactors in the future.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"44 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133434137","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
R. Cai, N. Yue, Hongyu Fang, Baowen Chen, Lili Liu, Zehua Ma
The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.
{"title":"Effects of Rolling on Characteristics of System Under Forced Circulation and Natural Circulation","authors":"R. Cai, N. Yue, Hongyu Fang, Baowen Chen, Lili Liu, Zehua Ma","doi":"10.1115/icone2020-16164","DOIUrl":"https://doi.org/10.1115/icone2020-16164","url":null,"abstract":"\u0000 The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117132501","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A Very High Temperature Reactor (VHTR) is one of the next generation nuclear systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. We measured the outlet flow rate by hot-wire anemometer which is an omnidirectional spherical probe of diameter 2.5mm. The experiment has been carried out under the condition that a copper wire with a scourer model and a cubic lattice model were inserting into the channel.
{"title":"Heat Transfer and Fluid Flow Characteristic of One Side Heated Vertical Rectangular Channel That Inserted Thin Metallic Wire","authors":"Gota Suga, T. Takeda","doi":"10.1115/icone2020-16705","DOIUrl":"https://doi.org/10.1115/icone2020-16705","url":null,"abstract":"\u0000 A Very High Temperature Reactor (VHTR) is one of the next generation nuclear systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. We measured the outlet flow rate by hot-wire anemometer which is an omnidirectional spherical probe of diameter 2.5mm. The experiment has been carried out under the condition that a copper wire with a scourer model and a cubic lattice model were inserting into the channel.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130093037","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Swirling flow is one of the well-recognized techniques to control the working process. This special flow is widely adopted in swirl vane separators in nuclear steam generator (SG) for water droplet separation and the fission gas removal system in Thorium Molten Salt Reactor (TMSR) for gas bubble separation. Since the parameters such as separation efficiency, pressure drop and mass and heat transfer rate are strongly dependent on the flow pattern, the accurate prediction of flow patterns and their transitions is extremely important for the proper design, operation and optimization of swirling two-phase flow systems. In this paper, using air and water as working fluids, a visualization experiment is carried out to study the gas-liquid flow in a horizontal pipe containing a swirler with four helical vanes. The test pipe is 5 m in length and 30 mm in diameter. Firstly, five typical flow patterns of swirling gas-liquid flow at the outlet of the swirler are classified and defined, these being spiral chain, swirling gas column, swirling intermittent, swirling annular and swirling ribbon flow. Being affected by the different gas and liquid flow rate of non-swirling flow, it is found that the same non-swirling flow can change into different swirling flow patterns. After that, the evolution of various swirling flow patterns along the streamwise direction is analyzed considering the influence of swirl attenuation. The results indicate that the same swirling flow pattern can transform into a variety of swirling flow patterns and subsequent non-swirling flow patterns. Finally, the flow pattern maps at different positions downstream of the swirler are presented.
{"title":"Experimental Study on Flow Patterns of Decaying Swirling Gas-Liquid Flow in a Horizontal Pipe","authors":"L. Shuai, Li Liu, Jiarong Zhang, Gu Hanyang","doi":"10.1115/icone2020-16224","DOIUrl":"https://doi.org/10.1115/icone2020-16224","url":null,"abstract":"\u0000 Swirling flow is one of the well-recognized techniques to control the working process. This special flow is widely adopted in swirl vane separators in nuclear steam generator (SG) for water droplet separation and the fission gas removal system in Thorium Molten Salt Reactor (TMSR) for gas bubble separation. Since the parameters such as separation efficiency, pressure drop and mass and heat transfer rate are strongly dependent on the flow pattern, the accurate prediction of flow patterns and their transitions is extremely important for the proper design, operation and optimization of swirling two-phase flow systems. In this paper, using air and water as working fluids, a visualization experiment is carried out to study the gas-liquid flow in a horizontal pipe containing a swirler with four helical vanes. The test pipe is 5 m in length and 30 mm in diameter. Firstly, five typical flow patterns of swirling gas-liquid flow at the outlet of the swirler are classified and defined, these being spiral chain, swirling gas column, swirling intermittent, swirling annular and swirling ribbon flow. Being affected by the different gas and liquid flow rate of non-swirling flow, it is found that the same non-swirling flow can change into different swirling flow patterns. After that, the evolution of various swirling flow patterns along the streamwise direction is analyzed considering the influence of swirl attenuation. The results indicate that the same swirling flow pattern can transform into a variety of swirling flow patterns and subsequent non-swirling flow patterns. Finally, the flow pattern maps at different positions downstream of the swirler are presented.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"121 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116046711","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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