Qingyang Zhang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang, Xukai Fan
This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.
本文开发了一个多物理接口程序MC-FLUENT，将蒙特卡罗程序OpenMC与商业计算流体动力学程序ANSYS FLUENT相耦合。研究了块高斯-塞德尔型和块雅可比型Picard迭代算法的实现及其并行性能。此外，本文还将两种自适应负载平衡算法引入到中子学和热工水力学耦合仿真中，以降低计算的时间成本。考虑到OpenMC和FLUENT的不同可扩展性限制了块高斯-塞德尔算法的性能，提出了一种可以动态增加节点数量的自适应负载平衡算法来提高其效率。此外，利用块雅可比算法的自然并行性，提出了另一种自适应负载均衡算法来提高其性能。A 3 x 3压水堆燃料引脚模型和1000 使用MWt-ABR金属基准核对两种算法的性能进行了比较，验证了两种自适应负载均衡算法的有效性。结果表明，当节点数量较大时，本文提出的自适应负载均衡算法可以大大提高块Jacobi算法的计算效率，并提高块Gauss–Seidel算法的性能。此外，当一个案例需要不同的OpenMC和FLUENT计算能力时，自适应负载平衡算法尤其有效。
{"title":"An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing","authors":"Qingyang Zhang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang, Xukai Fan","doi":"10.1155/2021/5549602","DOIUrl":"https://doi.org/10.1155/2021/5549602","url":null,"abstract":"This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48446795","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Libing Zhu, Jianxun Zhao, Xincheng Xiang, Yu Zhou, Xiangang Wang
The geometrical shape of the TRISO-coated particle is closely related to its performance and safety. In this paper, models were set up to study the failure fraction of TRISO particle, considering the real asphericity induced by manufacturing uncertainties. TRISO is simplified as a pressure vessel model, and micro X-ray CT was employed to detect the real geometrical shape. Key geometrical parameters, thickness and volume of the real particle, were then obtained with the 3D measurement method and input into PANAMA code (a German code for fuel performance simulation). Release fraction of fission gas and failure fraction of the TRISO-coated particle were revised with the aforementioned parameters with more accuracy and compared with those of the spherical particle. Obvious increment of failure fraction of the particle is found, which may contribute to the release of fission products.
{"title":"Failure Fraction Calculation of the TRISO-Coated Particle Using X-Ray Computed Tomography","authors":"Libing Zhu, Jianxun Zhao, Xincheng Xiang, Yu Zhou, Xiangang Wang","doi":"10.1155/2021/2420526","DOIUrl":"https://doi.org/10.1155/2021/2420526","url":null,"abstract":"The geometrical shape of the TRISO-coated particle is closely related to its performance and safety. In this paper, models were set up to study the failure fraction of TRISO particle, considering the real asphericity induced by manufacturing uncertainties. TRISO is simplified as a pressure vessel model, and micro X-ray CT was employed to detect the real geometrical shape. Key geometrical parameters, thickness and volume of the real particle, were then obtained with the 3D measurement method and input into PANAMA code (a German code for fuel performance simulation). Release fraction of fission gas and failure fraction of the TRISO-coated particle were revised with the aforementioned parameters with more accuracy and compared with those of the spherical particle. Obvious increment of failure fraction of the particle is found, which may contribute to the release of fission products.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41718282","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Tiejun Zeng, Xiaohua Yang, Yaping Wan, Panpan Jiang, L. Zhenghai
The loss and theft of radioactive material in transport can be attributed to the illegal movement. In order to distinguish it from the movement caused by the turbulence of the transportation vehicle, this paper proposes the criterion of “illegal movement” as the movement of radioactive materials outside the transportation compartment. Since the interior of the compartment is generally a metal environment, this paper proposes wireless signal strength data as a sensing method. The wireless signal strength data is filtered and converted into distance data. We construct a spatial triangle perpendicular to the top and sides of the compartment based on the distance data. When the radioactive material is inside the compartment, the angle between its corresponding point and the top plane of the compartment is less than 90°. Once it moves out of the compartment, the angle will be greater than 90°. Based on this, a sensing method of “illegal movement” based on spatial triangles is proposed. The simulation research shows that the scheme proposed in this paper is feasible.
{"title":"Detection of Illegal Movement in Radioactive Material Transportation Security Systems","authors":"Tiejun Zeng, Xiaohua Yang, Yaping Wan, Panpan Jiang, L. Zhenghai","doi":"10.1155/2021/6670114","DOIUrl":"https://doi.org/10.1155/2021/6670114","url":null,"abstract":"The loss and theft of radioactive material in transport can be attributed to the illegal movement. In order to distinguish it from the movement caused by the turbulence of the transportation vehicle, this paper proposes the criterion of “illegal movement” as the movement of radioactive materials outside the transportation compartment. Since the interior of the compartment is generally a metal environment, this paper proposes wireless signal strength data as a sensing method. The wireless signal strength data is filtered and converted into distance data. We construct a spatial triangle perpendicular to the top and sides of the compartment based on the distance data. When the radioactive material is inside the compartment, the angle between its corresponding point and the top plane of the compartment is less than 90°. Once it moves out of the compartment, the angle will be greater than 90°. Based on this, a sensing method of “illegal movement” based on spatial triangles is proposed. The simulation research shows that the scheme proposed in this paper is feasible.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47612955","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.
{"title":"Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method","authors":"M. Annor-Nyarko, Hong Xia","doi":"10.1155/2021/7035787","DOIUrl":"https://doi.org/10.1155/2021/7035787","url":null,"abstract":"The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49555313","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Zhen-Ying Wang, Zhiyun Liu, T. Ma, Chen Sun, Liu Liu, Yu Huang, Gang-Miao Guo
In order to improve the capability of dealing with software common cause failure (CCF) of digital reactor protection and monitoring system (RPMS), the diverse actuation system (DAS) is introduced for ACPR1000 nuclear power plants. From economic and feasibility point of view, the solution of DAS sharing with RPMS sensors and actuators is suggested; after capturing the function requirement of DAS, the automatic functions and manual functions assigned to it are determined based on transient analysis of design basic accidents concurrent with software CCF of RPMS. The independent verification proves that the reactor can be fallen back to and maintained at safety shutdown state, thanks to these DAS functions. Insight into probabilistic safety assessment proves significant reductions of risks are contributed. The critical technical issues while implementing DAS, such as measures to ensure its diversity from RPMS, precautions for preventing from its spurious actuation, isolation and independency from RPMS, and its testability and maintainability, are deliberately settled to improve its engineering reliability and alleviate the impact on RPMS as far as possible. Field programmable gate array technology that is diversified from RPMS is chosen to build DAS of ACPR1000 nuclear power plant, and the commissioning test verifies that it is capable of performing its designed functions. At last, a set of DAS-specific, paper-based, and event-oriented emergency operating procedure is developed, verified, and validated. Until now, the DAS system has always been successfully operating in all ACPR1000 nuclear power plants for several years.
{"title":"The Implementation of Diverse Actuation System in ACPR1000 Nuclear Power Plants","authors":"Zhen-Ying Wang, Zhiyun Liu, T. Ma, Chen Sun, Liu Liu, Yu Huang, Gang-Miao Guo","doi":"10.1155/2021/5529570","DOIUrl":"https://doi.org/10.1155/2021/5529570","url":null,"abstract":"In order to improve the capability of dealing with software common cause failure (CCF) of digital reactor protection and monitoring system (RPMS), the diverse actuation system (DAS) is introduced for ACPR1000 nuclear power plants. From economic and feasibility point of view, the solution of DAS sharing with RPMS sensors and actuators is suggested; after capturing the function requirement of DAS, the automatic functions and manual functions assigned to it are determined based on transient analysis of design basic accidents concurrent with software CCF of RPMS. The independent verification proves that the reactor can be fallen back to and maintained at safety shutdown state, thanks to these DAS functions. Insight into probabilistic safety assessment proves significant reductions of risks are contributed. The critical technical issues while implementing DAS, such as measures to ensure its diversity from RPMS, precautions for preventing from its spurious actuation, isolation and independency from RPMS, and its testability and maintainability, are deliberately settled to improve its engineering reliability and alleviate the impact on RPMS as far as possible. Field programmable gate array technology that is diversified from RPMS is chosen to build DAS of ACPR1000 nuclear power plant, and the commissioning test verifies that it is capable of performing its designed functions. At last, a set of DAS-specific, paper-based, and event-oriented emergency operating procedure is developed, verified, and validated. Until now, the DAS system has always been successfully operating in all ACPR1000 nuclear power plants for several years.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46297297","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.
{"title":"Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident","authors":"Shuliang Zou, Na Liu, B. Huang","doi":"10.1155/2021/1299821","DOIUrl":"https://doi.org/10.1155/2021/1299821","url":null,"abstract":"Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"1 1","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41374830","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Gas-cooled space nuclear reactor system usually utilizes the helium-xenon gas mixture as the working fluid. Since the typical helium-xenon mixture has the Prandtl number of about 0.2, which is lower than that of water and air, the turbulent flow and heat transfer features need to be further investigated among the helium-xenon mixture and other fluids. In the current paper, numerical investigations by ANSYS Fluent are performed on helium-xenon mixture flow (HeXe40, M = 40.0 g/mol, Pr = 0.21), airflow (Pr = 0.71), and water flow (Pr = 6.99) in the circular tube. Direct numerical simulation results of liquid metal flow (Pr = 0.01) are also adopted for comparison. Results show that the dimensionless velocity profile and shear stress in the boundary layer of HeXe40 are close to those of other fluids. The empirical correlations from other fluids can also predict well the friction factor of helium-xenon mixtures. Due to the discrepancy in turbulent heat diffusivity ratio, the dimensionless radial temperature profile and turbulent heat conduction of HeXe40 significantly differ from those of other fluids. The molecular conduction region of HeXe40 develops up to y+ ≈ 30 and extends to the logarithmic region of the flow boundary layer. Moreover, the available experimental Nusselt numbers of helium-xenon mixtures are compared with several convective heat transfer correlations, in which Kays correlation is better.
{"title":"Numerical Investigation on Turbulent Flow and Heat Transfer of Helium-Xenon Gas Mixture in a Circular Tube","authors":"Biao Zhou, Han Zhang, Yu Ji, Jun Sun, Yuliang Sun","doi":"10.1155/2021/8356893","DOIUrl":"https://doi.org/10.1155/2021/8356893","url":null,"abstract":"Gas-cooled space nuclear reactor system usually utilizes the helium-xenon gas mixture as the working fluid. Since the typical helium-xenon mixture has the Prandtl number of about 0.2, which is lower than that of water and air, the turbulent flow and heat transfer features need to be further investigated among the helium-xenon mixture and other fluids. In the current paper, numerical investigations by ANSYS Fluent are performed on helium-xenon mixture flow (HeXe40, M = 40.0 g/mol, Pr = 0.21), airflow (Pr = 0.71), and water flow (Pr = 6.99) in the circular tube. Direct numerical simulation results of liquid metal flow (Pr = 0.01) are also adopted for comparison. Results show that the dimensionless velocity profile and shear stress in the boundary layer of HeXe40 are close to those of other fluids. The empirical correlations from other fluids can also predict well the friction factor of helium-xenon mixtures. Due to the discrepancy in turbulent heat diffusivity ratio, the dimensionless radial temperature profile and turbulent heat conduction of HeXe40 significantly differ from those of other fluids. The molecular conduction region of HeXe40 develops up to y+ ≈ 30 and extends to the logarithmic region of the flow boundary layer. Moreover, the available experimental Nusselt numbers of helium-xenon mixtures are compared with several convective heat transfer correlations, in which Kays correlation is better.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47775712","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Antai Liu, Chang-qi Yan, Fuqiang Zhu, Haifeng Gu, Suijun Gong
As two important parameters, the velocity of disturbance wave and the wall shear stress in annular flow are very important to solve the closed equations of the mechanical model for annular flow. In this study, the disturbance wave velocity and wall shear stress of annular flow in a vertical narrow rectangular channel with a cross section of 70 mm × 2 mm were studied. According to the experimental results, it is found that the wave velocity and wall shear stress of disturbance wave increase with increasing gas phase velocity and liquid phase velocity. Also, existing correlations for predicting the velocity of disturbance wave were summarized and evaluated using the current experimental data. A new correlation for wall shear stress based on the disturbance wave velocity has been proposed. Compared with the existing correlation for predicting wall shear stress, this new correlation can well predict the current experimental data and MAPE is only 7.32%.
扰动波速度和壁面剪切应力作为环空流动的两个重要参数，对于求解环空流动力学模型的封闭方程是非常重要的。本文研究了横断面为70 mm × 2 mm的垂直窄矩形通道中环形流动的扰动波速和壁面剪应力。实验结果表明，扰动波的波速和壁面剪应力随气相速度和液相速度的增大而增大。同时，利用现有的实验数据对已有的干扰波速度预测相关性进行了总结和评价。提出了一种新的基于扰动波速的壁面剪应力关系式。与已有的墙体剪应力预测相关性相比，该相关性能较好地预测当前实验数据，MAPE仅为7.32%。
{"title":"Studying Disturbance Wave Velocity and Wall Shear Stress of Vertical Upward Annular Flow in Narrow Rectangular Channel","authors":"Antai Liu, Chang-qi Yan, Fuqiang Zhu, Haifeng Gu, Suijun Gong","doi":"10.1155/2021/1927479","DOIUrl":"https://doi.org/10.1155/2021/1927479","url":null,"abstract":"As two important parameters, the velocity of disturbance wave and the wall shear stress in annular flow are very important to solve the closed equations of the mechanical model for annular flow. In this study, the disturbance wave velocity and wall shear stress of annular flow in a vertical narrow rectangular channel with a cross section of 70 mm × 2 mm were studied. According to the experimental results, it is found that the wave velocity and wall shear stress of disturbance wave increase with increasing gas phase velocity and liquid phase velocity. Also, existing correlations for predicting the velocity of disturbance wave were summarized and evaluated using the current experimental data. A new correlation for wall shear stress based on the disturbance wave velocity has been proposed. Compared with the existing correlation for predicting wall shear stress, this new correlation can well predict the current experimental data and MAPE is only 7.32%.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42136060","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nuclear power plant operating data are characterized by a large variety, strong coupling, and low data value density. When using machine learning techniques for fault diagnosis and other related research, feature selection enables dimensionality reduction while maintaining the physical meaning of the original features, thus improving the computational efficiency and generalization ability of the learning model. In this paper, a correlation-based feature selection algorithm is developed to implement feature selection of nuclear power plant operating data. The proposed algorithm is verified by experiments and compared with traditional correlation-based feature selection algorithms. The experiments and comparison results show that the proposed algorithm is effective in realizing the dimensionality reduction of nuclear power plant operating data.
{"title":"A Correlation-Based Feature Selection Algorithm for Operating Data of Nuclear Power Plants","authors":"Yuxuan He, Hongxing Yu, Ren Yu, Jian Song, Haibo Lian, Jiangyang He, Jiangtao Yuan","doi":"10.1155/2021/9994340","DOIUrl":"https://doi.org/10.1155/2021/9994340","url":null,"abstract":"Nuclear power plant operating data are characterized by a large variety, strong coupling, and low data value density. When using machine learning techniques for fault diagnosis and other related research, feature selection enables dimensionality reduction while maintaining the physical meaning of the original features, thus improving the computational efficiency and generalization ability of the learning model. In this paper, a correlation-based feature selection algorithm is developed to implement feature selection of nuclear power plant operating data. The proposed algorithm is verified by experiments and compared with traditional correlation-based feature selection algorithms. The experiments and comparison results show that the proposed algorithm is effective in realizing the dimensionality reduction of nuclear power plant operating data.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46628899","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The experimental facility VULCAN was setup to study the fuel-coolant interaction (FCI) phenomena in a postulated severe accident of light water reactors. The heating system is important for the facility to prepare molten material in a crucible. This article is concerned with the design of the heating system, which applies electromagnetic induction heating method. The COMSOL code was employed to simulate the induction heating characteristics of a graphite crucible under different current and frequency of the work coil (inductor). Given a frequency, the relationship between the crucible’s average temperature and the inductor’s current is obtained, which is instrumental to select the power supply of the induction heating system. Meanwhile, the skin effect of induction heating is analyzed to guide the choice of frequency and inductor of the heating system. According to the simulation results, the induction heating system of frequency 47 kHz is suitable for the experiment, with a good agreement in temperature between the measured and the predicted.
{"title":"COMSOL Simulation for Design of Induction Heating System in VULCAN Facility","authors":"Jinkun Min, Guangyu Zhu, Yi Yuan, Jingquan Liu","doi":"10.1155/2021/9922503","DOIUrl":"https://doi.org/10.1155/2021/9922503","url":null,"abstract":"The experimental facility VULCAN was setup to study the fuel-coolant interaction (FCI) phenomena in a postulated severe accident of light water reactors. The heating system is important for the facility to prepare molten material in a crucible. This article is concerned with the design of the heating system, which applies electromagnetic induction heating method. The COMSOL code was employed to simulate the induction heating characteristics of a graphite crucible under different current and frequency of the work coil (inductor). Given a frequency, the relationship between the crucible’s average temperature and the inductor’s current is obtained, which is instrumental to select the power supply of the induction heating system. Meanwhile, the skin effect of induction heating is analyzed to guide the choice of frequency and inductor of the heating system. According to the simulation results, the induction heating system of frequency 47 kHz is suitable for the experiment, with a good agreement in temperature between the measured and the predicted.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-08-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45176433","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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