Jung-Hoon Choi, Byeonggwan Lee, Ki-rak Lee, H. Kang, H. J. Eom, Seong-Sik Shin, Ga-Yeong Kim, Hwan-seo Park
To reduce the environmental burden caused by the disposal of spent nuclear fuel, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. To manufacture a waste form of high durability, the characteristics of the waste generated during the process should be evaluated. In this study, the physical, radiological, and thermal characteristics of the waste and waste forms for major nuclides (Cs, Sr, I, transuranic/rare earth, and Tc/Se) generated in the nuclide management process were analyzed. In the case of Cs nuclides, characterization was conducted according to the capture rate of the adsorbent in the high-temperature heat treatment process; meanwhile, in the case of Sr nuclides, characterization was performed by considering the ratio of similar nuclides in the chlorination process. For I nuclide, analysis was performed based on the available waste form, and for TRU/RE and Tc/Se nuclides, analysis was performed by considering chlorination and mid-temperature heat treatment. The radioactivity and heat generation rate of each waste and waste form were evaluated over a period of 1,000 years. The results of this study could be used to derive the centerline temperature for the thermal stability evaluation of waste forms and for the feasibility evaluation of each disposal system considered in the waste burden minimization technology.
{"title":"Characterization of Waste Generated from Nuclide Management Process in Waste Burden Minimization Technology for Spent Nuclear Fuel","authors":"Jung-Hoon Choi, Byeonggwan Lee, Ki-rak Lee, H. Kang, H. J. Eom, Seong-Sik Shin, Ga-Yeong Kim, Hwan-seo Park","doi":"10.1155/2022/4764825","DOIUrl":"https://doi.org/10.1155/2022/4764825","url":null,"abstract":"To reduce the environmental burden caused by the disposal of spent nuclear fuel, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. To manufacture a waste form of high durability, the characteristics of the waste generated during the process should be evaluated. In this study, the physical, radiological, and thermal characteristics of the waste and waste forms for major nuclides (Cs, Sr, I, transuranic/rare earth, and Tc/Se) generated in the nuclide management process were analyzed. In the case of Cs nuclides, characterization was conducted according to the capture rate of the adsorbent in the high-temperature heat treatment process; meanwhile, in the case of Sr nuclides, characterization was performed by considering the ratio of similar nuclides in the chlorination process. For I nuclide, analysis was performed based on the available waste form, and for TRU/RE and Tc/Se nuclides, analysis was performed by considering chlorination and mid-temperature heat treatment. The radioactivity and heat generation rate of each waste and waste form were evaluated over a period of 1,000 years. The results of this study could be used to derive the centerline temperature for the thermal stability evaluation of waste forms and for the feasibility evaluation of each disposal system considered in the waste burden minimization technology.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48172368","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}
S. Pinem, S. Dibyo, W. Luthfi, V. Wardhani, D. Hartanto
Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.
{"title":"An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes","authors":"S. Pinem, S. Dibyo, W. Luthfi, V. Wardhani, D. Hartanto","doi":"10.1155/2022/6030504","DOIUrl":"https://doi.org/10.1155/2022/6030504","url":null,"abstract":"Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49574823","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}
Gaojun Liu, Weijie Fan, Feng-lei Li, Gaixia Wang, Dongdong You
Aiming at solving the problems of small fault data samples and insufficient remaining useful life (RUL) prediction accuracy of nuclear power machinery, a method based on an exponential degradation model is proposed to predict the RUL of equipment after the failure warning system alarm. After data preprocessing, time-domain feature extraction, selection, and dimensionality reduction fusion of multiple degradation variables, the exponential degradation model is constructed based on the Bayesian process, and prior information is used. As an application, the RUL of a nuclear power turbine was calculated based on actual monitoring data, the α − λ precision curve was used to evaluate the prediction effect, and the RUL prediction results verified the effectiveness of the proposed method.
{"title":"Remaining Useful Life Prediction of Nuclear Power Machinery Based on an Exponential Degradation Model","authors":"Gaojun Liu, Weijie Fan, Feng-lei Li, Gaixia Wang, Dongdong You","doi":"10.1155/2022/9895907","DOIUrl":"https://doi.org/10.1155/2022/9895907","url":null,"abstract":"Aiming at solving the problems of small fault data samples and insufficient remaining useful life (RUL) prediction accuracy of nuclear power machinery, a method based on an exponential degradation model is proposed to predict the RUL of equipment after the failure warning system alarm. After data preprocessing, time-domain feature extraction, selection, and dimensionality reduction fusion of multiple degradation variables, the exponential degradation model is constructed based on the Bayesian process, and prior information is used. As an application, the RUL of a nuclear power turbine was calculated based on actual monitoring data, the \u0000 \u0000 α\u0000 −\u0000 λ\u0000 \u0000 precision curve was used to evaluate the prediction effect, and the RUL prediction results verified the effectiveness of the proposed method.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48007101","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 radionuclide dispersion in coastal water is mainly controlled by the water flow and tidal effect. Tracing and analysis of radioactive pollutant dispersion in coastal water can predict distribution of radionuclide under marine transportation accident of spent fuel. In this work, factors such as continuous emission, radioactive decay, and water depth are considered, and a hydrodynamic model of radionuclide dispersion based on shallow water equations is established to simulate the dispersion of the radioactive pollutant in coastal waters under different hydrological conditions. As far as the characteristics of the radionuclide dispersion in coastal water are concerned, the simulation of pollutants by the hydrodynamic model is in good agreement with the work of Bailly du Bois et al., which validated the correctness of this model. The model has been applied to simulate the distribution of radionuclides in coastal water following a marine transport accident of spent fuel near Daya Bay Nuclear Power Plant in China. The simulation reveals that the distribution features are significantly affected by different hydrological conditions. In addition to limiting the diffusion range, the vortex effect can also cause the accumulation of radionuclides near the vortex, which helps to provide more practical information for nuclear emergency decision makers.
放射性核素在沿海水体中的扩散主要受水流和潮汐效应的控制。对放射性污染物在近海的扩散进行跟踪分析，可以预测乏燃料海上运输事故下放射性核素的分布。本文考虑了连续排放、放射性衰变和水深等因素，建立了基于浅水方程的放射性核素扩散水动力学模型，模拟了不同水文条件下放射性污染物在沿海水域的扩散。就放射性核素在沿海水中的扩散特性而言，流体动力学模型对污染物的模拟与Bailly du Bois等人的工作非常一致，验证了该模型的正确性。该模型已用于模拟中国大亚湾核电站附近乏燃料海上运输事故后沿海水中放射性核素的分布。模拟结果表明，不同水文条件对其分布特征有显著影响。涡流效应除了限制扩散范围外，还可以导致放射性核素在涡流附近积聚，这有助于为核应急决策者提供更实用的信息。
{"title":"Study on the Dispersion of Radionuclides under Different Hydrological Conditions of Spent Fuel Shipping in Daya Bay","authors":"Liwei Chen, Wei Chen, Jiazhen Lin, Chunhua Chen, Yalin Luo, Longlong Tao","doi":"10.1155/2022/7265821","DOIUrl":"https://doi.org/10.1155/2022/7265821","url":null,"abstract":"The radionuclide dispersion in coastal water is mainly controlled by the water flow and tidal effect. Tracing and analysis of radioactive pollutant dispersion in coastal water can predict distribution of radionuclide under marine transportation accident of spent fuel. In this work, factors such as continuous emission, radioactive decay, and water depth are considered, and a hydrodynamic model of radionuclide dispersion based on shallow water equations is established to simulate the dispersion of the radioactive pollutant in coastal waters under different hydrological conditions. As far as the characteristics of the radionuclide dispersion in coastal water are concerned, the simulation of pollutants by the hydrodynamic model is in good agreement with the work of Bailly du Bois et al., which validated the correctness of this model. The model has been applied to simulate the distribution of radionuclides in coastal water following a marine transport accident of spent fuel near Daya Bay Nuclear Power Plant in China. The simulation reveals that the distribution features are significantly affected by different hydrological conditions. In addition to limiting the diffusion range, the vortex effect can also cause the accumulation of radionuclides near the vortex, which helps to provide more practical information for nuclear emergency decision makers.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-06-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49473921","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}
Fang Liu, Yuanyuan Zhou, Wencheng Song, Hongzhi Wang
This study aimed to investigate the inhibitory effects of cold atmospheric plasma (CAP) on anaplastic thyroid cancer cells (CAL-62 cells) and to reveal the molecular mechanism. The effects of CAP on CAL-62 cells were evaluated by cell viability, superoxide dismutase activity, apoptosis, cell cycle, and protein expression level, and the role of reactive oxygen species (ROS) produced by plasma was also investigated. The results showed that CAP dose-dependently inhibited cell viability and promotes cell apoptosis and G2/M arrest by increasing cell ROS levels. The activity of superoxide dismutase (SOD) was enhanced by CAP which indicated that the antioxidant system of the cell was activated. Additionally, the ROS produced by CAP can inhibit CAL-62 cell proliferation by inhibiting the PI3K/Akt/mTOR signaling pathway. Therefore, these findings will provide useful support for the application of CAP for treating anaplastic thyroid cancer.
{"title":"Cold Atmospheric Plasma Inhibits the Proliferation of CAL-62 Cells through the ROS-Mediated PI3K/Akt/mTOR Signaling Pathway","authors":"Fang Liu, Yuanyuan Zhou, Wencheng Song, Hongzhi Wang","doi":"10.1155/2022/3884695","DOIUrl":"https://doi.org/10.1155/2022/3884695","url":null,"abstract":"This study aimed to investigate the inhibitory effects of cold atmospheric plasma (CAP) on anaplastic thyroid cancer cells (CAL-62 cells) and to reveal the molecular mechanism. The effects of CAP on CAL-62 cells were evaluated by cell viability, superoxide dismutase activity, apoptosis, cell cycle, and protein expression level, and the role of reactive oxygen species (ROS) produced by plasma was also investigated. The results showed that CAP dose-dependently inhibited cell viability and promotes cell apoptosis and G2/M arrest by increasing cell ROS levels. The activity of superoxide dismutase (SOD) was enhanced by CAP which indicated that the antioxidant system of the cell was activated. Additionally, the ROS produced by CAP can inhibit CAL-62 cell proliferation by inhibiting the PI3K/Akt/mTOR signaling pathway. Therefore, these findings will provide useful support for the application of CAP for treating anaplastic thyroid cancer.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-06-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48734796","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 main transformer is critical equipment for economically generating electricity in nuclear power plants (NPPs). Dissolved gas analysis (DGA) is an effective means of monitoring the transformer condition, and its parameters can reflect the transformer operating condition. This study introduces a framework for main transformer predictive-based maintenance management. A condition prediction method based on the online support vector machine (SVM) regression model is proposed, with the input data being preprocessed using the information granulation method, and the parameters of the model are optimized using the particle swarm optimization (PSO) algorithm. Using DGA data from the NPP data acquisition system, two experiments are designed to verify the trend tracing and prediction envelope ability of main transformers installed in NPPs with different operating ages of the proposed model. Finally, how to use this framework to benefit the maintenance plan of the main transformer is summarized.
{"title":"An Information Granulated Based SVM Approach for Anomaly Detection of Main Transformers in Nuclear Power Plants","authors":"Wenmin Yu, Ren Yu, Cheng Li","doi":"10.1155/2022/3931374","DOIUrl":"https://doi.org/10.1155/2022/3931374","url":null,"abstract":"The main transformer is critical equipment for economically generating electricity in nuclear power plants (NPPs). Dissolved gas analysis (DGA) is an effective means of monitoring the transformer condition, and its parameters can reflect the transformer operating condition. This study introduces a framework for main transformer predictive-based maintenance management. A condition prediction method based on the online support vector machine (SVM) regression model is proposed, with the input data being preprocessed using the information granulation method, and the parameters of the model are optimized using the particle swarm optimization (PSO) algorithm. Using DGA data from the NPP data acquisition system, two experiments are designed to verify the trend tracing and prediction envelope ability of main transformers installed in NPPs with different operating ages of the proposed model. Finally, how to use this framework to benefit the maintenance plan of the main transformer is summarized.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-06-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46517841","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 plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.
{"title":"Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis","authors":"M. Zubair","doi":"10.1155/2022/4666161","DOIUrl":"https://doi.org/10.1155/2022/4666161","url":null,"abstract":"Nuclear power plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45154876","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}
Fangfang Zhang, Nana Li, Di Zhu, R. Xiao, Weichao Liu, R. Tao
In general, weak compressibility is one of the properties of liquids. That is, in actual operation of hydraulic machinery, the flow is weakly compressible. However, the influence of weak compressibility is often neglected in usual numerical simulation, which makes the simulation results different from the experimental results. Based on the Computational Fluid Dynamics (CFD) solver and model test rig, by means of mutual verification between numerical simulation and experiment, the fitting degree between numerical results and experimental results before and after considering weak compressibility is compared and analyzed in this paper; it is obtained that the numerical results is closer to the experimental results after considering the weak compressibility. In addition, velocity field of pump turbines, head loss of main components, and the change of entropy yield are analyzed and reasons for numerical value being closer to the experimental value after considering weak compressibility of fluid are summarized and analyzed. It is proved that the consideration of weak compressibility is of great significance to improve the accuracy of results in the numerical simulation of pump turbines.
{"title":"Influence of Weak Compressibility on the Hydrodynamic Performance Evaluation of Pump Turbines in the Pump Mode","authors":"Fangfang Zhang, Nana Li, Di Zhu, R. Xiao, Weichao Liu, R. Tao","doi":"10.1155/2022/3544436","DOIUrl":"https://doi.org/10.1155/2022/3544436","url":null,"abstract":"In general, weak compressibility is one of the properties of liquids. That is, in actual operation of hydraulic machinery, the flow is weakly compressible. However, the influence of weak compressibility is often neglected in usual numerical simulation, which makes the simulation results different from the experimental results. Based on the Computational Fluid Dynamics (CFD) solver and model test rig, by means of mutual verification between numerical simulation and experiment, the fitting degree between numerical results and experimental results before and after considering weak compressibility is compared and analyzed in this paper; it is obtained that the numerical results is closer to the experimental results after considering the weak compressibility. In addition, velocity field of pump turbines, head loss of main components, and the change of entropy yield are analyzed and reasons for numerical value being closer to the experimental value after considering weak compressibility of fluid are summarized and analyzed. It is proved that the consideration of weak compressibility is of great significance to improve the accuracy of results in the numerical simulation of pump turbines.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-05-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42854701","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}
Because of the characteristics of the small water volume of OTSG, it is hard to control the outlet steam pressure when the load is changed or disturbed. This study is devoted to the control of the once-through steam generator (OTSG). A double-layer controller based on the PPO algorithm is proposed to control the outlet steam pressure of OTSG. The bottom layer is the PID controller; it directly regulates the OTSG feed water valve and then controls the steam pressure. The top layer of the controller is the agent based on the PPO algorithm, which is responsible for optimizing the parameters of the PID in real time to obtain better control performance. The agent chooses PID parameters as actions to the environment, and then, the reward value is obtained through the reward function of the environment which enables online learning of the agent. Compared with the PID controller, the simulation experiment result shows that the method not only has a good control performance but also has a good anti-interference ability.
{"title":"Design of Control System of Once-Through Steam Generator Based on Proximal Policy Optimization Algorithm","authors":"Cheng Li, Ren Yu, Wenmin Yu, Tianshu Wang","doi":"10.1155/2022/2941705","DOIUrl":"https://doi.org/10.1155/2022/2941705","url":null,"abstract":"Because of the characteristics of the small water volume of OTSG, it is hard to control the outlet steam pressure when the load is changed or disturbed. This study is devoted to the control of the once-through steam generator (OTSG). A double-layer controller based on the PPO algorithm is proposed to control the outlet steam pressure of OTSG. The bottom layer is the PID controller; it directly regulates the OTSG feed water valve and then controls the steam pressure. The top layer of the controller is the agent based on the PPO algorithm, which is responsible for optimizing the parameters of the PID in real time to obtain better control performance. The agent chooses PID parameters as actions to the environment, and then, the reward value is obtained through the reward function of the environment which enables online learning of the agent. Compared with the PID controller, the simulation experiment result shows that the method not only has a good control performance but also has a good anti-interference ability.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-05-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44359845","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}
Yu‐Liang Zhang, Jin-Fu Li, Tao Wang, Jun-Jian Xiao, Xiaoqi Jia, Li Zhang
In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.
{"title":"Pressure Distribution on the Inner Wall of the Volute Casing of a Centrifugal Pump","authors":"Yu‐Liang Zhang, Jin-Fu Li, Tao Wang, Jun-Jian Xiao, Xiaoqi Jia, Li Zhang","doi":"10.1155/2022/3563459","DOIUrl":"https://doi.org/10.1155/2022/3563459","url":null,"abstract":"In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-05-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43737775","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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