The blanket modules of first wall need bear tremendous heat flux due to the very high temperature of plasma in the nuclear fusion reactor. Therefore, it is significant to clarify the knowledge of transient heat transfer process for helium gas flowing in the tubes installed in the blanket modules. In this research, the transient heat transfer process of turbulent forced convection for helium gas flowing in a horizontal minichannel was experimentally investigated. The test tube made of platinum with the inner diameter of 1.8 mm, the wall thickness of 0.1 mm and the effective length of 90 mm was heated by a direct current from power source. The heat generation rate of the test tube, Q̇, was raised with an exponential function, Q̇ = Q0 exp(t/τ), where Q0 is the initial heat generation rate, t is time, and τ is e-folding time of heat generation rate. The heat generation rates of the test tube were controlled and measured by a heat input control system. The flow rates were adjusted by the bypass of gas loop and measured by the turbine flow meter. The experiment was conducted under the e-folding time of heat generation rate ranged from 40 ms to 15 s. Based on experimental data, it is obvious that the heat flux and temperature difference between surface temperature of test tube and bulk temperature of helium gas increased with the exponentially increasing of heat generation rate. At the same flow velocity, the heat transfer coefficients approached constant values when the e-folding time is longer than about 1 s (quasi-steady state), but increased with a decrease of e-folding time when the e-folding time is smaller than about 1 s (transient state). The heat transfer coefficients increased with the increase in flow velocities but showed less dependent on flow velocities at shorter e-folding time. Furthermore, the Nusselt number under quasi-steady and transient condition was affected by the Reynolds number and the Fourier number.
{"title":"Experimental Study on Transient Heat Transfer for Helium Gas Flowing in a Minichannel","authors":"Feng Xu, Qiusheng Liu, S. Kawaguchi, M. Shibahara","doi":"10.1115/icone2020-16697","DOIUrl":"https://doi.org/10.1115/icone2020-16697","url":null,"abstract":"\u0000 The blanket modules of first wall need bear tremendous heat flux due to the very high temperature of plasma in the nuclear fusion reactor. Therefore, it is significant to clarify the knowledge of transient heat transfer process for helium gas flowing in the tubes installed in the blanket modules.\u0000 In this research, the transient heat transfer process of turbulent forced convection for helium gas flowing in a horizontal minichannel was experimentally investigated. The test tube made of platinum with the inner diameter of 1.8 mm, the wall thickness of 0.1 mm and the effective length of 90 mm was heated by a direct current from power source. The heat generation rate of the test tube, Q̇, was raised with an exponential function, Q̇ = Q0 exp(t/τ), where Q0 is the initial heat generation rate, t is time, and τ is e-folding time of heat generation rate. The heat generation rates of the test tube were controlled and measured by a heat input control system. The flow rates were adjusted by the bypass of gas loop and measured by the turbine flow meter. The experiment was conducted under the e-folding time of heat generation rate ranged from 40 ms to 15 s. Based on experimental data, it is obvious that the heat flux and temperature difference between surface temperature of test tube and bulk temperature of helium gas increased with the exponentially increasing of heat generation rate. At the same flow velocity, the heat transfer coefficients approached constant values when the e-folding time is longer than about 1 s (quasi-steady state), but increased with a decrease of e-folding time when the e-folding time is smaller than about 1 s (transient state). The heat transfer coefficients increased with the increase in flow velocities but showed less dependent on flow velocities at shorter e-folding time. Furthermore, the Nusselt number under quasi-steady and transient condition was affected by the Reynolds number and the Fourier number.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"16 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":"126441176","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}
Linshan Wei, Shuhong Du, W. Gu, N. Zhang, M. Ding, Zhong-ning Sun, Zhaoming Meng
HPR1000 is an advanced nuclear power plant with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. It is based on the large accumulated knowledge from the design, construction as well as operations experience of nuclear power plants in China. The passive containment cooling system (PCS) of HPR1000 is an important and innovative passive safety system to suppress the pressure in the containment during LOCA. In this paper, the detailed design process of PCS is reviewed, and an integrated experiment facility for the study on the coupling behavior between PCS and thermal hydraulic characteristics in the containment is described, and arrangement of measuring points including temperature, pressure, gas composition and so on are introduced in detailed. Also, the experimental energy released and energy vent to ensure the similarity of containment pressure response, thermal stratification and PCS heat removal is introduced. According to this versatile experiment facility can conduct real-engineering system test which is designed to support the PCS development. In addition, this valuable experience in the design and manufacture of integrated experiment facility can provide important technical support and guidance for the China next generation advanced PWR as well as safety related system.
{"title":"An Integrated Experimental Test Facility to Support Development of the Passive Containment Cooling System of HPR1000","authors":"Linshan Wei, Shuhong Du, W. Gu, N. Zhang, M. Ding, Zhong-ning Sun, Zhaoming Meng","doi":"10.1115/icone2020-16678","DOIUrl":"https://doi.org/10.1115/icone2020-16678","url":null,"abstract":"\u0000 HPR1000 is an advanced nuclear power plant with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. It is based on the large accumulated knowledge from the design, construction as well as operations experience of nuclear power plants in China. The passive containment cooling system (PCS) of HPR1000 is an important and innovative passive safety system to suppress the pressure in the containment during LOCA. In this paper, the detailed design process of PCS is reviewed, and an integrated experiment facility for the study on the coupling behavior between PCS and thermal hydraulic characteristics in the containment is described, and arrangement of measuring points including temperature, pressure, gas composition and so on are introduced in detailed. Also, the experimental energy released and energy vent to ensure the similarity of containment pressure response, thermal stratification and PCS heat removal is introduced. According to this versatile experiment facility can conduct real-engineering system test which is designed to support the PCS development. In addition, this valuable experience in the design and manufacture of integrated experiment facility can provide important technical support and guidance for the China next generation advanced PWR as well as safety related system.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"89 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":"121310435","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 the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. However, the steam Direct Contact Condensation (DCC) microscopic mechanisms are very complicated, which are not very clear yet. Moreover, the high-pressure and high-temperature experiment is very expensive to be conducted for many different test conditions. So in the present work, both the experimental and numerical methods are employed to investigate the steam DCC behavior. The steam DCC experimental bench has been built up, and the key parameters including the flow patterns and steam core temperature distributions are measured to provide validation data for the numerical results. In aspect of the numerical work, CFD simulation on the steam condensation is conducted. The heat and mass transfer process is simulated through the three-dimension commercial software FLUENT 16.0. Some of the key heat and mass transfer correlations are added by User Defined Function (UDF). The key parameters including the condensation steam fraction, temperature, and pressure, etc. are analyzed, which reflect the major heat transfer characteristics. According to the results, the expansion-compression-steam tail could be observed in both the numerical and experimental results. In essential, the steam fraction, temperature, and pressure distributions are determined by the equilibrium and transformation between the thermal dynamic energy and kinetic energy. The results provide working references for the practical ADS steam spraying condensation process in AP1000 reactor.
{"title":"Experimental and Numerical Research on Steam Direct Contact Condensation Process in Automatic Depressurization System of AP1000","authors":"Yuhao Zhang, Li Feng, Z. Qiu, Jingpin Fu, D. Lu","doi":"10.1115/icone2020-16780","DOIUrl":"https://doi.org/10.1115/icone2020-16780","url":null,"abstract":"\u0000 In the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. However, the steam Direct Contact Condensation (DCC) microscopic mechanisms are very complicated, which are not very clear yet. Moreover, the high-pressure and high-temperature experiment is very expensive to be conducted for many different test conditions. So in the present work, both the experimental and numerical methods are employed to investigate the steam DCC behavior. The steam DCC experimental bench has been built up, and the key parameters including the flow patterns and steam core temperature distributions are measured to provide validation data for the numerical results. In aspect of the numerical work, CFD simulation on the steam condensation is conducted. The heat and mass transfer process is simulated through the three-dimension commercial software FLUENT 16.0. Some of the key heat and mass transfer correlations are added by User Defined Function (UDF). The key parameters including the condensation steam fraction, temperature, and pressure, etc. are analyzed, which reflect the major heat transfer characteristics. According to the results, the expansion-compression-steam tail could be observed in both the numerical and experimental results. In essential, the steam fraction, temperature, and pressure distributions are determined by the equilibrium and transformation between the thermal dynamic energy and kinetic energy. The results provide working references for the practical ADS steam spraying condensation process in AP1000 reactor.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"43 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":"123095953","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}
Anomaly detection is significant for the cybersecurity of the I&C systems at nuclear power plants. There are a large number of network packets generated in the network traffic of the I&C systems. There are many attributes of the network traffic can used for anomaly detection. The structure of the network packets is analyzed in detail with examples. Then, Features are extracted from network packets. An unsupervised neural network called autoencoder is applied for anomaly detection. Training and testing database are captured from a physical PLC system which simulates a water level control system. The result of the test results shows that the neural network can detect anomaly successfully.
{"title":"Anomaly Detection for Network Traffic of I&C Systems Based on Neural Network","authors":"Wen Si, Jianghai Li, Ronghong Qu, Xiaojin Huang","doi":"10.1115/icone2020-16900","DOIUrl":"https://doi.org/10.1115/icone2020-16900","url":null,"abstract":"\u0000 Anomaly detection is significant for the cybersecurity of the I&C systems at nuclear power plants. There are a large number of network packets generated in the network traffic of the I&C systems. There are many attributes of the network traffic can used for anomaly detection. The structure of the network packets is analyzed in detail with examples. Then, Features are extracted from network packets. An unsupervised neural network called autoencoder is applied for anomaly detection. Training and testing database are captured from a physical PLC system which simulates a water level control system. The result of the test results shows that the neural network can detect anomaly successfully.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"519 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":"115101809","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}
Pei Yu, Jiaming Wang, Huiyun Ma, Haifeng Gu, Chang-qi Yan
The steam hammer pressure is solved though the simplified calculation. PIPENET software is applied to model the nuclear island main steam system between the steam generator and the main steam header in HPR 1000. The transient module is used to simulate the occurrence and attenuation process of steam hammer. The maximum steam hammer pressure, the maximum steam hammer stress in the pipe system, when and where the load occurs are given. The influence of the straight pipe section length and valve closing time on the steam hammer effect is analyzed. With the other conditions unchanged, the steam hammer energy decreases as the straight pipe section shortens, or the valve closing time extends.
{"title":"Calculation and Analysis of Steam Hammer in Main Steam Pipe in HPR1000","authors":"Pei Yu, Jiaming Wang, Huiyun Ma, Haifeng Gu, Chang-qi Yan","doi":"10.1115/icone2020-17004","DOIUrl":"https://doi.org/10.1115/icone2020-17004","url":null,"abstract":"\u0000 The steam hammer pressure is solved though the simplified calculation. PIPENET software is applied to model the nuclear island main steam system between the steam generator and the main steam header in HPR 1000. The transient module is used to simulate the occurrence and attenuation process of steam hammer. The maximum steam hammer pressure, the maximum steam hammer stress in the pipe system, when and where the load occurs are given. The influence of the straight pipe section length and valve closing time on the steam hammer effect is analyzed. With the other conditions unchanged, the steam hammer energy decreases as the straight pipe section shortens, or the valve closing time extends.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"163 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":"133265456","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}
Quanyao Ren, Zeng Pu, M. Zheng, M. Su, Ping Chen, L. Pan, Hui He, Qingche He
The gas-liquid two-phase flow behaviors are always associated with its dynamic void fraction, such as flow resistance, heat transfer coefficient, phase distribution, critical heat flux etc. As regard to the commercial PWR and BWR, rod bundles are the typical geometry, which contains many sub-channels for coolant flowing. In present study, the sub-channel void fraction was measured in 5 × 5 rod bundles with the sub-channel impedance void meter consisting of 12 strip electrodes. Based on the measured void fraction in different sub-channels, the void fraction dynamics, PDF (probability distribution function) and CDF (cumulative distribution function) curves were analyzed to make clear the effect of superficial gas and liquid velocity, flow development and casing tube. The empirical correlation for PDF of dynamic sub-channel void fraction has been developed, which showed good fitness with PDF and CDF curves and satisfying accuracy of averaged void fraction.
{"title":"Experimental Study on the Sub-Channel Void Fraction Characteristics of Bubbly Flow in Rod Bundles","authors":"Quanyao Ren, Zeng Pu, M. Zheng, M. Su, Ping Chen, L. Pan, Hui He, Qingche He","doi":"10.1115/icone2020-16315","DOIUrl":"https://doi.org/10.1115/icone2020-16315","url":null,"abstract":"\u0000 The gas-liquid two-phase flow behaviors are always associated with its dynamic void fraction, such as flow resistance, heat transfer coefficient, phase distribution, critical heat flux etc. As regard to the commercial PWR and BWR, rod bundles are the typical geometry, which contains many sub-channels for coolant flowing. In present study, the sub-channel void fraction was measured in 5 × 5 rod bundles with the sub-channel impedance void meter consisting of 12 strip electrodes. Based on the measured void fraction in different sub-channels, the void fraction dynamics, PDF (probability distribution function) and CDF (cumulative distribution function) curves were analyzed to make clear the effect of superficial gas and liquid velocity, flow development and casing tube. The empirical correlation for PDF of dynamic sub-channel void fraction has been developed, which showed good fitness with PDF and CDF curves and satisfying accuracy of averaged void fraction.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"17 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":"130170190","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}
Wang Bolong, L. Weihua, Jia Haijun, L. Jun, Hao Wentao
Small reactors have received more and more attention for their high safety, reliability, low power density, and short construction period. And the gas-steam pressurizer is widely used in small reactors due to its characteristics of simple structure, saves the heating and spray equipment, and prevents the coolant from boiling. The gas-steam pressurizer is a pressure control equipment for the reactor coolant system, and its characteristic of transient response is an important factor that affect operation stability of nuclear reactor systems. An experimental system was established to study the effect of pressure response for an insurge transient and influence factors were analyzed quantitatively. Experimental investigation shows that for the gas-steam pressurizer, the increase of coolant loading capacity (insurge) can cause system pressure rising. And the change of system pressure has much consistency with the change of liquid level and gas space temperature. The liquid phase exists temperature fluctuations and overall shows a downward trend during the insurge transient. And there exists a temperature gradient from bottom to top in the pressurizer liquid phase region during the insurge transient. The change of water vapor quantity curve is the oscillating curve during the transient and water vapor quantity is in a decreasing trend overall during the insurge transient. What’s more, the experiments also analyzed the pressure response and temperature response during the insurge transient.
{"title":"Experimental Study of the Processes of Gas-Steam Pressurizer Insurge Transients","authors":"Wang Bolong, L. Weihua, Jia Haijun, L. Jun, Hao Wentao","doi":"10.1115/icone2020-16183","DOIUrl":"https://doi.org/10.1115/icone2020-16183","url":null,"abstract":"\u0000 Small reactors have received more and more attention for their high safety, reliability, low power density, and short construction period. And the gas-steam pressurizer is widely used in small reactors due to its characteristics of simple structure, saves the heating and spray equipment, and prevents the coolant from boiling. The gas-steam pressurizer is a pressure control equipment for the reactor coolant system, and its characteristic of transient response is an important factor that affect operation stability of nuclear reactor systems. An experimental system was established to study the effect of pressure response for an insurge transient and influence factors were analyzed quantitatively. Experimental investigation shows that for the gas-steam pressurizer, the increase of coolant loading capacity (insurge) can cause system pressure rising. And the change of system pressure has much consistency with the change of liquid level and gas space temperature. The liquid phase exists temperature fluctuations and overall shows a downward trend during the insurge transient. And there exists a temperature gradient from bottom to top in the pressurizer liquid phase region during the insurge transient. The change of water vapor quantity curve is the oscillating curve during the transient and water vapor quantity is in a decreasing trend overall during the insurge transient. What’s more, the experiments also analyzed the pressure response and temperature response during the insurge transient.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"29 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":"114557896","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}
The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.
{"title":"Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor","authors":"G. Hawkes","doi":"10.1115/icone2020-16592","DOIUrl":"https://doi.org/10.1115/icone2020-16592","url":null,"abstract":"\u0000 The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated.\u0000 Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"77 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":"134609835","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}
Bowen Chen, Bo Wang, Bingzheng Ke, Ru Li, Ruifeng Tian
The steam generator is an important part of the nuclear power plant, and the corrugated plate separator plays the important role of drying steam in the steam generator to improve power generation efficiency and protect the safety of the system. The separation mechanism of the corrugated plate separator is relatively complicated. The droplets are moved by the drag force of the steam and gravity in the corrugated plate separator, and captured by the wall of the corrugated plate separator. When the velocity is increased, the inertial force of droplet is increased, so that the droplet is more easily captured by the wall of the corrugated plate separator, and the separation efficiency of the corrugated plate separator is increased. In this paper, the phenomenon of droplet impact on the inclined wall is studied by high-speed photography technology, and the cause and mechanism of the phenomenon are analyzed. By analyzing the spreading and splashing on the droplets impacting on the inclined wall, the relationship between the inclination angle of the droplet impacting on the inclined wall and the spreading is obtained, and the influence of droplets with different Weber numbers, and dry and wetted walls were analyzed, which provide a basis for the optimization of the corrugated plate separator.
{"title":"Analysis of Droplet Impacting on Inclined Wall","authors":"Bowen Chen, Bo Wang, Bingzheng Ke, Ru Li, Ruifeng Tian","doi":"10.1115/icone2020-16964","DOIUrl":"https://doi.org/10.1115/icone2020-16964","url":null,"abstract":"\u0000 The steam generator is an important part of the nuclear power plant, and the corrugated plate separator plays the important role of drying steam in the steam generator to improve power generation efficiency and protect the safety of the system. The separation mechanism of the corrugated plate separator is relatively complicated. The droplets are moved by the drag force of the steam and gravity in the corrugated plate separator, and captured by the wall of the corrugated plate separator. When the velocity is increased, the inertial force of droplet is increased, so that the droplet is more easily captured by the wall of the corrugated plate separator, and the separation efficiency of the corrugated plate separator is increased. In this paper, the phenomenon of droplet impact on the inclined wall is studied by high-speed photography technology, and the cause and mechanism of the phenomenon are analyzed. By analyzing the spreading and splashing on the droplets impacting on the inclined wall, the relationship between the inclination angle of the droplet impacting on the inclined wall and the spreading is obtained, and the influence of droplets with different Weber numbers, and dry and wetted walls were analyzed, which provide a basis for the optimization of the corrugated plate separator.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"41 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":"132795880","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 lot of previous works indicated that the 3D neutronics calculation module can provide better prediction results in reactor system safety analysis especially under several anticipated transient accidents which with strong spatial influence factors. SAC-3D code is a new version of SAC series reactor system safety analysis codes which was developed by North China Electric Power University[4][5][6][7]. In this version, the neutronics calculation module was developed based on the three-dimensional nodal expansion method. In 2012, IAEA initialed a CRP on benchmark analysis of EBR-II loss of flow without scram test, SHRT-45R benchmark data is one of the benchmark specifications provided by ANL. In the present work, the neutronics benchmark problem of EBR-II SHRT-45R was analyzed with SAC-3D. The neutronics calculation module of SAC-3D was validated by comparing the key results with the benchmark data. The simulation results also agreed well with the results provided by other participants of this CRP.
{"title":"Code to Code Validation of SAC-3D Based on EBR-II Benchmark Problem","authors":"D. Lu, Lyu Siyu, D. Sui","doi":"10.1115/icone2020-16666","DOIUrl":"https://doi.org/10.1115/icone2020-16666","url":null,"abstract":"\u0000 A lot of previous works indicated that the 3D neutronics calculation module can provide better prediction results in reactor system safety analysis especially under several anticipated transient accidents which with strong spatial influence factors. SAC-3D code is a new version of SAC series reactor system safety analysis codes which was developed by North China Electric Power University[4][5][6][7]. In this version, the neutronics calculation module was developed based on the three-dimensional nodal expansion method. In 2012, IAEA initialed a CRP on benchmark analysis of EBR-II loss of flow without scram test, SHRT-45R benchmark data is one of the benchmark specifications provided by ANL. In the present work, the neutronics benchmark problem of EBR-II SHRT-45R was analyzed with SAC-3D. The neutronics calculation module of SAC-3D was validated by comparing the key results with the benchmark data. The simulation results also agreed well with the results provided by other participants of this CRP.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"31 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":"124716651","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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