The effect, the pebble spacing has, on the fluid flow past two spheres in a free stream, was investigated by utilizing Direct Numerical Simulation (DNS), a Spectral Element Method (SEM) and NEK5000. There was seen that for a spacing of 2 pebble diameters (D), between the centroids, the turbulence statistics were asymmetric. For this study four different pebble spacings were investigated: 1.5D, 2.0D, 2.5LD, 3.0D. For all the cases the inlet flow, (U0) was selected that the Reynolds number, (Re) equalled 1000. The flow was allowed to develop for 200 Convective Units (CU = D/U0) before tests were conducted to ensure compliance to DNS. With the Kolmogorov scales captured and adequate Y+ values, time averaging was done for 800 CU to calculate the turbulence statistics. Calculated parameters include turbulence statists for selected line profiles and turbulent kinetic energy (TKE) budget terms. The generated statistics will be compared for each spacing selection.
{"title":"The Influence of Pebble Placement on the Wake of Tandem Pebbles in a Free Stream","authors":"Gerrit Botha, Y. Hassan, R. Kurwitz, E. Merzari","doi":"10.1115/ICONE26-81884","DOIUrl":"https://doi.org/10.1115/ICONE26-81884","url":null,"abstract":"The effect, the pebble spacing has, on the fluid flow past two spheres in a free stream, was investigated by utilizing Direct Numerical Simulation (DNS), a Spectral Element Method (SEM) and NEK5000. There was seen that for a spacing of 2 pebble diameters (D), between the centroids, the turbulence statistics were asymmetric. For this study four different pebble spacings were investigated: 1.5D, 2.0D, 2.5LD, 3.0D. For all the cases the inlet flow, (U0) was selected that the Reynolds number, (Re) equalled 1000. The flow was allowed to develop for 200 Convective Units (CU = D/U0) before tests were conducted to ensure compliance to DNS. With the Kolmogorov scales captured and adequate Y+ values, time averaging was done for 800 CU to calculate the turbulence statistics. Calculated parameters include turbulence statists for selected line profiles and turbulent kinetic energy (TKE) budget terms. The generated statistics will be compared for each spacing selection.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"33 5 Pt A 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133383424","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}
J. Chen, Dalin Zhang, S. Qiu, Zhang Kui, Mingjun Wang, G. Su
As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.
{"title":"CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor","authors":"J. Chen, Dalin Zhang, S. Qiu, Zhang Kui, Mingjun Wang, G. Su","doi":"10.1115/ICONE26-81626","DOIUrl":"https://doi.org/10.1115/ICONE26-81626","url":null,"abstract":"As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"62 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131992928","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}
Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.
蒸汽喷射器作为应对核反应堆严重事故的一种手段，越来越受到人们的关注。它是一种静压喷射泵，利用驱动力将蒸汽和水两种流体通过直接接触冷凝而降低内部压力。此外，还对SI作为高传热换热器的性能进行了展望。没有驱动单元(如外部电源和旋转机器)是SI的重要特征，可以预期它会降低安装和维护成本。也有可能产生高于进口压力的排放压力。因此，即使在核电站断电的情况下，SI也可以作为冷却堆芯的静态安全系统使用。虽然SI很早以前就被用于蒸汽机，但其运行机制尚未阐明。因此，阐明SI的运行机制对于核电站的介绍是必不可少的。建立了一个一维分析模型，该模型可以预测在完全冷凝和评估排放压力的情况下的运行特性(Narabayashi等人，1996)。此外，通过详细观察，证实了扩散器截面存在亮度边界(Abe et al.， 2012)。这被认为是两相流凝结的边界。然而，目前的分析模型并没有考虑到这种现象。本研究的目的是澄清扩散器部分的流动结构。为此，利用高速摄像机观察透明SI测试件扩散器截面的状态，并同时测量其各点的压力。当背压阀关闭时，确认亮度边界接近喉部。除了这个边界外，还证实了明亮区域是间歇性向下游传播的。假定这种现象是由压力增加引起的，并假定其传播为以声速运动的压力波。因此，通过计算图像处理的传播速度来估计空隙率。在此基础上，分别采用大、中、小三种试验件进行了试验。从以上结果出发，讨论了SI扩压器截面内部流动结构。
{"title":"Study on Flow Structure in a Supersonic Steam Injector","authors":"Yuki Kamata, Masaya Fujishiro, A. Kaneko, Y. Abe","doi":"10.1115/ICONE26-82058","DOIUrl":"https://doi.org/10.1115/ICONE26-82058","url":null,"abstract":"Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant.\u0000 Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model.\u0000 The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114162949","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}
After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the potential for recriticality of the fuel debris, which is formed after the core meltdown. In this study, a conservative criticality evaluation of the Fukushima Daiichi Unit 1 debris bed has been carried out. Parameters, such as debris size, porosity, particle size, fuel burnup and the coolant conditions, including the water density and the content of boron, were considered. The effect of these parameters on the neutron multiplication factor was analysed and safety parameter ranges, i.e. zones where the recriticality can be totally excluded, have been identified. The content of boron in water required to secure the subcriticality was calculated for those zones with recriticality potential. It was found that recriticality is achievable for a wide range of fuel debris conditions. 1600 ppm B would ensure subcriticality under any conditions.
{"title":"A Criticality Evaluation of Fukushima Daiichi Unit 1 Fuel Debris","authors":"M. F. López, M. Buck, J. Starflinger","doi":"10.1115/ICONE26-81148","DOIUrl":"https://doi.org/10.1115/ICONE26-81148","url":null,"abstract":"After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the potential for recriticality of the fuel debris, which is formed after the core meltdown. In this study, a conservative criticality evaluation of the Fukushima Daiichi Unit 1 debris bed has been carried out. Parameters, such as debris size, porosity, particle size, fuel burnup and the coolant conditions, including the water density and the content of boron, were considered. The effect of these parameters on the neutron multiplication factor was analysed and safety parameter ranges, i.e. zones where the recriticality can be totally excluded, have been identified. The content of boron in water required to secure the subcriticality was calculated for those zones with recriticality potential. It was found that recriticality is achievable for a wide range of fuel debris conditions. 1600 ppm B would ensure subcriticality under any conditions.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114899411","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 Chong, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Sun Guanghui
In recent years, nuclear power are widely used in the ship and submarine. Affected by the ocean wave, the bubble dynamic under ocean conditions may be different from stationary condition. In order to investigate the influence mechanism of ocean conditions, the single bubble rising in stagnant liquid under rolling conditions with different frequencies and amplitudes were calculated through CFD method. In present work, the typical ocean conditions such as rolling were realized through dynamic mesh method, this method could simulated the movement of container directly. The Volume-of-fluid (VOF) method was applied to track the interface between liquid and gas phases and the surface tension were calculated by Continuum Surface Force (CSF) method. We can draw the conclusions as follows: (1) the bubble tend to move laterally and periodically under rolling conditions, this may attributed to the additional force caused by rolling motion. (2) the period of lateral movement are in accordance with the rolling period, and the frequency and amplitude of bubble lateral movement may increasing as the decreasing of rolling period. (3) The amplitude of the bubble lateral displacement is proportional to the rolling amplitude. (4)larger bubbles have the higher rising velocity than the small ones, and the larger bubble are easier to break up which may promote the lateral movement.
{"title":"Numerical Simulation of Bubble Dynamic Under Ocean Conditions","authors":"Chen Chong, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Sun Guanghui","doi":"10.1115/ICONE26-81639","DOIUrl":"https://doi.org/10.1115/ICONE26-81639","url":null,"abstract":"In recent years, nuclear power are widely used in the ship and submarine. Affected by the ocean wave, the bubble dynamic under ocean conditions may be different from stationary condition. In order to investigate the influence mechanism of ocean conditions, the single bubble rising in stagnant liquid under rolling conditions with different frequencies and amplitudes were calculated through CFD method. In present work, the typical ocean conditions such as rolling were realized through dynamic mesh method, this method could simulated the movement of container directly. The Volume-of-fluid (VOF) method was applied to track the interface between liquid and gas phases and the surface tension were calculated by Continuum Surface Force (CSF) method. We can draw the conclusions as follows: (1) the bubble tend to move laterally and periodically under rolling conditions, this may attributed to the additional force caused by rolling motion. (2) the period of lateral movement are in accordance with the rolling period, and the frequency and amplitude of bubble lateral movement may increasing as the decreasing of rolling period. (3) The amplitude of the bubble lateral displacement is proportional to the rolling amplitude. (4)larger bubbles have the higher rising velocity than the small ones, and the larger bubble are easier to break up which may promote the lateral movement.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"12 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125376873","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}
Bian Jiawei, Dalin Zhang, Rulei Sun, Yingwei Wu, W. Tian, G. Su, S. Qiu
Spraying system plays an important role in the safety of PWR. To ensure homogeneous spraying of the containment, the layout of nozzles on the spray header was taken into consideration in design. In this paper, an experimental study was conducted to obtain spray characteristics data, including spray cone angle and 2-D spray flux distribution for the purpose of achieving optimal design of the spraying system. According to the specialty of the spray field involved, a testing loop with four pressure-swirl nozzles was established for the study. Spray cone angles were obtained by photograph method. The volume flux distribution was measured by collecting the spray droplet along the cross-section diameters. Based on the experimental data, typical spray flux distributions were obtained. The flux distribution results were used to build 3-D coverage models. Then these models were used to calculate the overall spray coverage in the containment. The present work introduces the experimental study of spray behavior of a typical pressure-swirl nozzle in containment and the method to evaluate spray coverage through building 3-D spray flux distribution models. The work is expected to be helpful for the optimization design of spraying systems.
{"title":"Experimental Study on Spray Pattern of Pressure-Swirl Nozzle in Reactor Containment","authors":"Bian Jiawei, Dalin Zhang, Rulei Sun, Yingwei Wu, W. Tian, G. Su, S. Qiu","doi":"10.1115/ICONE26-81505","DOIUrl":"https://doi.org/10.1115/ICONE26-81505","url":null,"abstract":"Spraying system plays an important role in the safety of PWR. To ensure homogeneous spraying of the containment, the layout of nozzles on the spray header was taken into consideration in design. In this paper, an experimental study was conducted to obtain spray characteristics data, including spray cone angle and 2-D spray flux distribution for the purpose of achieving optimal design of the spraying system.\u0000 According to the specialty of the spray field involved, a testing loop with four pressure-swirl nozzles was established for the study. Spray cone angles were obtained by photograph method. The volume flux distribution was measured by collecting the spray droplet along the cross-section diameters. Based on the experimental data, typical spray flux distributions were obtained. The flux distribution results were used to build 3-D coverage models. Then these models were used to calculate the overall spray coverage in the containment. The present work introduces the experimental study of spray behavior of a typical pressure-swirl nozzle in containment and the method to evaluate spray coverage through building 3-D spray flux distribution models. The work is expected to be helpful for the optimization design of spraying systems.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"35 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121904089","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}
When considering the safety of the reactor after the core melted, the external reactor vessel cooling (IVR-ERVC) is regarded as one of the most prominent method and is now widely being studied. And in order to apply this method more efficiently, CHF is the utmost part because it limits the upper threshold of the cooling effect. There has already been a large number of investigation on the CHF both by experiment and simulation. But for the experiment, most of them used the copper for its high thermal conductivity. However, the lower plenum is actually composed of the carbon steel instead of copper since the reactor pressure vessel and some core catchers in nuclear power plants are made of this material. This CHF experiment here used carbon steel plate on a downward face. The carbon steel plate was attached to the copper base by solder. The results of the carbon steel plate experiment were then analyzed. After polishing the surface by sandpapers, the heat flux is being increased until CHF was reached. We found that the surface is somehow changed during the heating process. This process was repeated several times, and the growing tendency of the CHF was found. Detailed images of the heating surface acquired by high speed camera under different heat fluxes were also obtained and analyzed. It was found that the more the oxidization was, the fewer bubbles were generated and the bigger the CHF was. Finally some theories of the CHF mechanism were also analyzed. The authors hope that study could shed some light on surface effect on causing different CHF.
{"title":"Preliminary Research on the Oxidation Effect of the Carbon Steel Plate of Downward Facing Pool Boiling by Two-Dimensional Image","authors":"Kai Wang, N. Erkan, K. Okamoto","doi":"10.1115/ICONE26-82019","DOIUrl":"https://doi.org/10.1115/ICONE26-82019","url":null,"abstract":"When considering the safety of the reactor after the core melted, the external reactor vessel cooling (IVR-ERVC) is regarded as one of the most prominent method and is now widely being studied. And in order to apply this method more efficiently, CHF is the utmost part because it limits the upper threshold of the cooling effect. There has already been a large number of investigation on the CHF both by experiment and simulation. But for the experiment, most of them used the copper for its high thermal conductivity. However, the lower plenum is actually composed of the carbon steel instead of copper since the reactor pressure vessel and some core catchers in nuclear power plants are made of this material.\u0000 This CHF experiment here used carbon steel plate on a downward face. The carbon steel plate was attached to the copper base by solder. The results of the carbon steel plate experiment were then analyzed. After polishing the surface by sandpapers, the heat flux is being increased until CHF was reached. We found that the surface is somehow changed during the heating process. This process was repeated several times, and the growing tendency of the CHF was found. Detailed images of the heating surface acquired by high speed camera under different heat fluxes were also obtained and analyzed. It was found that the more the oxidization was, the fewer bubbles were generated and the bigger the CHF was. Finally some theories of the CHF mechanism were also analyzed. The authors hope that study could shed some light on surface effect on causing different CHF.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"38 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117164621","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}
T. Sugimoto, Shimpei Saito, A. Kaneko, Y. Abe, A. Uchibori, H. Ohshima
A sodium-cooled fast reactor (SFR) is now under development in Japan. A shell-and-tube type once-through heat exchanger is to be installed to generate steam in the design. Low-pressure hot sodium flows in the shell side and high-pressure water, which heated to become steam, flows in the tube side. It has been anticipated that a pin hole is formed on the tube wall and high-pressure steam blows out from the hole. When a high-pressure steam flows out from the tube hole, a high-speed steam jet is formed in the sodium coolant. Fine sodium droplets are torn off from the sodium surface and entrained into the steam jet. Sodium-water chemical reaction causes an increase of entrained droplet temperature. The hot and high-speed sodium entrained droplets attack the wall of a neighboring tube and cause a wastage on the tube wall, which may lead to a failure propagation. In Japan Atomic Energy Agency (JAEA), an analysis code for the sodium-water reaction phenomenon, called SERAPHIM, has already been developed. Visualization data is required to validate the liquid entrainment model in this code. Since the flow velocity at the gas leakage is a sonic speed, it is extremely difficult to visualize the inside of the gas jet. Experiments have been carried out to visualize this phenomenon in the past; however, experimental data for model validation has not been entirely obtained due to the above-mentioned difficulty. Thus, the motivation of this study is to examine the possibility of visualization method and to clarify flow structure. To this end, we first performed the preliminary experiments using simple test facilities. Two types of test sections were used in the experiments: three-dimensional one and two-dimensional one. In the experiment using the three-dimensional one, we tried to visualize a more realistic phenomenon. Through this experiment, the whole gas-jet behavior was clearly captured. However, we found that the detailed droplet-entrainment behavior in a gas jet could not be obtained in this setup, especially at high-velocity conditions. Then, we carried out the experiments using the two-dimensional one. In these experiments, the flow structure of a gas jet was simplified. However, it was difficult to distinguish the liquid film formed on the wall surface of the test section from the entrained droplets. We considered that the liquid film is formed due to the nozzle outlet shape and improved the test section. By experiments with new test section, we succeeded in visualizing entrained droplets of relatively large diameter and calculated droplet diameter distribution. Then, we discussed the mechanism of entrained droplet behavior.
{"title":"Visualization Study on Droplet-Entrainment in a High-Speed Gas Jet Into a Liquid Pool","authors":"T. Sugimoto, Shimpei Saito, A. Kaneko, Y. Abe, A. Uchibori, H. Ohshima","doi":"10.1115/ICONE26-81695","DOIUrl":"https://doi.org/10.1115/ICONE26-81695","url":null,"abstract":"A sodium-cooled fast reactor (SFR) is now under development in Japan. A shell-and-tube type once-through heat exchanger is to be installed to generate steam in the design. Low-pressure hot sodium flows in the shell side and high-pressure water, which heated to become steam, flows in the tube side. It has been anticipated that a pin hole is formed on the tube wall and high-pressure steam blows out from the hole. When a high-pressure steam flows out from the tube hole, a high-speed steam jet is formed in the sodium coolant. Fine sodium droplets are torn off from the sodium surface and entrained into the steam jet. Sodium-water chemical reaction causes an increase of entrained droplet temperature. The hot and high-speed sodium entrained droplets attack the wall of a neighboring tube and cause a wastage on the tube wall, which may lead to a failure propagation.\u0000 In Japan Atomic Energy Agency (JAEA), an analysis code for the sodium-water reaction phenomenon, called SERAPHIM, has already been developed. Visualization data is required to validate the liquid entrainment model in this code. Since the flow velocity at the gas leakage is a sonic speed, it is extremely difficult to visualize the inside of the gas jet. Experiments have been carried out to visualize this phenomenon in the past; however, experimental data for model validation has not been entirely obtained due to the above-mentioned difficulty. Thus, the motivation of this study is to examine the possibility of visualization method and to clarify flow structure.\u0000 To this end, we first performed the preliminary experiments using simple test facilities. Two types of test sections were used in the experiments: three-dimensional one and two-dimensional one. In the experiment using the three-dimensional one, we tried to visualize a more realistic phenomenon. Through this experiment, the whole gas-jet behavior was clearly captured. However, we found that the detailed droplet-entrainment behavior in a gas jet could not be obtained in this setup, especially at high-velocity conditions. Then, we carried out the experiments using the two-dimensional one. In these experiments, the flow structure of a gas jet was simplified. However, it was difficult to distinguish the liquid film formed on the wall surface of the test section from the entrained droplets. We considered that the liquid film is formed due to the nozzle outlet shape and improved the test section. By experiments with new test section, we succeeded in visualizing entrained droplets of relatively large diameter and calculated droplet diameter distribution. Then, we discussed the mechanism of entrained droplet behavior.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130968043","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}
With the purpose of enhancement of heat transfer performance and reduction of the volume of steam generator (SG), a structure of longitudinal finned tubes was proposed to replace the smooth tubes of SG in this paper. Taking the SG smooth tubes of Daya bay Nuclear Power plant as a reference, the simplified heat transfer model of new longitudinal finned tubes was established by ANSYS CFX. Three-dimensional numerical model was developed to investigate the fluid-solid coupled thermal hydraulic characteristics of different types of the longitudinal finned tubes compared with the smooth tubes. Analysis of calculation results were sufficiently discussed for the effect of mass flow rate, fin array, solid thermal conductivity and frictional resistance. The numerical results revealed that the heat transfer coefficient increase with the increasing mass flow rate in the secondary side. The material of the tubes has significantly influence on the heat transfer process. Different flow conditions have different thermal hydraulic characteristics. The evaluated criterion to judge the enhancement of the heat transfer of the coupled process was also proposed. The numerical results can provide some useful guidance for design optimization of longitudinal finned tubes in SG.
{"title":"Numerical Investigation on the Heat Transfer Enhancement Behavior Outside Longitudinal Finned Tubes","authors":"Yujia Zhou, H. Bo, Jingyu Du","doi":"10.1115/ICONE26-81283","DOIUrl":"https://doi.org/10.1115/ICONE26-81283","url":null,"abstract":"With the purpose of enhancement of heat transfer performance and reduction of the volume of steam generator (SG), a structure of longitudinal finned tubes was proposed to replace the smooth tubes of SG in this paper. Taking the SG smooth tubes of Daya bay Nuclear Power plant as a reference, the simplified heat transfer model of new longitudinal finned tubes was established by ANSYS CFX. Three-dimensional numerical model was developed to investigate the fluid-solid coupled thermal hydraulic characteristics of different types of the longitudinal finned tubes compared with the smooth tubes. Analysis of calculation results were sufficiently discussed for the effect of mass flow rate, fin array, solid thermal conductivity and frictional resistance. The numerical results revealed that the heat transfer coefficient increase with the increasing mass flow rate in the secondary side. The material of the tubes has significantly influence on the heat transfer process. Different flow conditions have different thermal hydraulic characteristics. The evaluated criterion to judge the enhancement of the heat transfer of the coupled process was also proposed. The numerical results can provide some useful guidance for design optimization of longitudinal finned tubes in SG.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"37 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129735033","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}
In a 2008 report on safety analysis for research reactors, the International Atomic Energy Agency (IAEA) identified experimentation as the preferred method of code validation [1]. However, many experiments currently used for code validation are performed under conditions that are not representative of real nuclear systems. Furthermore, the predominant uncertainties reported for reactor systems parameters are typically those associated with evaluated nuclear data libraries however, the significance of spatial uncertainties remains generally unknown. The magnitude of local flux measurement experimental uncertainties have not be investigated at length in the McMaster Nuclear Reactor (MNR). Such results can be used for validation of MNR models with both Monte Carlo N Particle (MCNP) and Serpent code packages. Flux wire measurements have previously been conducted at the center of an irradiation site (via the technique of neutron activation analysis), where a locally uniform flux distribution has been assumed. Early stage results show good agreement with three-dimensional neutron diffusion theory and demonstrate the viability of such measurements for continued analysis. However, the magnitude of the effects of Xenon buildup, control rod positions, and spatial sample positioning on the data remain unknown, and so a series of experiments is ongoing to address these areas of experimental variability. Full length flux wire irradiations at several high-power levels (500 kW, 800 kW, and 1 MW) are being conducted to quantify these effects. At each operating power level, several NiCr wires are irradiated, and the decay of 51Cr examined to determine the total neutron flux in the irradiation site. The use of multiple wires per irradiation provides insight into the spatial gradient of the neutron flux across one reactor site (approximately 8 × 8 cm).
{"title":"Axial Flux Wire Measurements at the McMaster Nuclear Reactor","authors":"E. MacConnachie, D. Novog, S. Day","doi":"10.1115/ICONE26-82412","DOIUrl":"https://doi.org/10.1115/ICONE26-82412","url":null,"abstract":"In a 2008 report on safety analysis for research reactors, the International Atomic Energy Agency (IAEA) identified experimentation as the preferred method of code validation [1]. However, many experiments currently used for code validation are performed under conditions that are not representative of real nuclear systems. Furthermore, the predominant uncertainties reported for reactor systems parameters are typically those associated with evaluated nuclear data libraries however, the significance of spatial uncertainties remains generally unknown. The magnitude of local flux measurement experimental uncertainties have not be investigated at length in the McMaster Nuclear Reactor (MNR). Such results can be used for validation of MNR models with both Monte Carlo N Particle (MCNP) and Serpent code packages. Flux wire measurements have previously been conducted at the center of an irradiation site (via the technique of neutron activation analysis), where a locally uniform flux distribution has been assumed. Early stage results show good agreement with three-dimensional neutron diffusion theory and demonstrate the viability of such measurements for continued analysis. However, the magnitude of the effects of Xenon buildup, control rod positions, and spatial sample positioning on the data remain unknown, and so a series of experiments is ongoing to address these areas of experimental variability. Full length flux wire irradiations at several high-power levels (500 kW, 800 kW, and 1 MW) are being conducted to quantify these effects. At each operating power level, several NiCr wires are irradiated, and the decay of 51Cr examined to determine the total neutron flux in the irradiation site. The use of multiple wires per irradiation provides insight into the spatial gradient of the neutron flux across one reactor site (approximately 8 × 8 cm).","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"40 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128875534","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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