The development of nuclear energy will play an important role in the implementation of China’s carbon peaking and carbon neutral strategy. Since China’s first nuclear power plant came into commercial operation, nuclear power generation has reached a total of more than 2.6 trillion kilowatt hours, equivalent to a reduction of carbon dioxide emissions of about 2.1 billion tons. With more nuclear power units coming into operation, the production of spent fuel has also increased exponentially, and spent fuel transportation and off-stack storage have become the most urgent aspects of the back-end of China’s nuclear fuel cycle. The need for large spent fuel transport containers is becoming increasingly urgent as the spent fuel pools in China are becoming full and spent fuel cooled to a certain age needs to be transported to the final disposal site in spent fuel transport containers. The paper investigates the current research status of spent fuel transport containers from three aspects: design and manufacture, test capability and numerical simulation analysis, and gives reasonable suggestions for the development of large spent fuel transport containers in China.
{"title":"Research Status of Spent Fuel Transport Containers","authors":"Changwu Wang, Zhipeng Wang, Qian Sun, Yuhang Zhang","doi":"10.1115/icone29-93580","DOIUrl":"https://doi.org/10.1115/icone29-93580","url":null,"abstract":"\u0000 The development of nuclear energy will play an important role in the implementation of China’s carbon peaking and carbon neutral strategy. Since China’s first nuclear power plant came into commercial operation, nuclear power generation has reached a total of more than 2.6 trillion kilowatt hours, equivalent to a reduction of carbon dioxide emissions of about 2.1 billion tons. With more nuclear power units coming into operation, the production of spent fuel has also increased exponentially, and spent fuel transportation and off-stack storage have become the most urgent aspects of the back-end of China’s nuclear fuel cycle. The need for large spent fuel transport containers is becoming increasingly urgent as the spent fuel pools in China are becoming full and spent fuel cooled to a certain age needs to be transported to the final disposal site in spent fuel transport containers. The paper investigates the current research status of spent fuel transport containers from three aspects: design and manufacture, test capability and numerical simulation analysis, and gives reasonable suggestions for the development of large spent fuel transport containers in China.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116559396","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-Vessel Retention (IVR) strategy is one of the most important severe accident management strategies. The Critical Heat Flux (CHF) is an important key element in the IVR research. How to increase the CHF get more and more attention in the CHF research. This paper aims at the influence of the coolant with different impurity. Those impurity can fall into the coolant when the severe accident happens. In this experiment, the deionized water was used as the coolant. The experiment investigated the different impurity with boronic acid that has different concentration and the boronic acid water with different impurity, such as concrete, fiberglass and coating. The concrete can be fallen from the construction, the fiberglass can be fallen from the insulation cotton which cover the pipe and device, and the coating can be fallen from the devices and other constructions. When the accident happens, those impurity can fall from the construction, then fall into the coolant and influence the CHF of the outside surface of lower head. From the result of the experiment, the CHF with the boron acid water is higher than that only deionized water. The CHF with the impurity water depend on the concentration of the impurity. Lower concentration will increase the CHF when the higher concentration will decrease the CHF. Different impurity has different influence on CHF, we control the component of the coolant to improve the CHF.
{"title":"The Influence of Impurity Water for CHF on the Downward Facing Boiling Surface","authors":"Bo-Yu Lin, Xiangyu Yun, Junying Xu, Dongshan Wei, Huiyong Zhang","doi":"10.1115/icone29-93157","DOIUrl":"https://doi.org/10.1115/icone29-93157","url":null,"abstract":"\u0000 In-Vessel Retention (IVR) strategy is one of the most important severe accident management strategies. The Critical Heat Flux (CHF) is an important key element in the IVR research. How to increase the CHF get more and more attention in the CHF research. This paper aims at the influence of the coolant with different impurity. Those impurity can fall into the coolant when the severe accident happens. In this experiment, the deionized water was used as the coolant. The experiment investigated the different impurity with boronic acid that has different concentration and the boronic acid water with different impurity, such as concrete, fiberglass and coating. The concrete can be fallen from the construction, the fiberglass can be fallen from the insulation cotton which cover the pipe and device, and the coating can be fallen from the devices and other constructions. When the accident happens, those impurity can fall from the construction, then fall into the coolant and influence the CHF of the outside surface of lower head. From the result of the experiment, the CHF with the boron acid water is higher than that only deionized water. The CHF with the impurity water depend on the concentration of the impurity. Lower concentration will increase the CHF when the higher concentration will decrease the CHF. Different impurity has different influence on CHF, we control the component of the coolant to improve the CHF.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"7 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117263507","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}
There have been heated discussions on how to reduce the earthquake response of NPPs. Generally, isolation technology is one of the most efficient methods but researches on traditional isolated structure and 3D isolated structure for NPPs, to some extent, have deficiencies. Therefore, in this paper, a new structural system which included base isolation system and lateral damping supporters for NPPs (hereinafter referred to as DISO) in a cave or underground was proposed to obtain a positive seismic response, aiming to protect the nuclear island. The motion characteristics of the system have been studied and a simplified mathematical model has been set up. The dynamic equation was deduced and solved. The theoretical analysis results that there is an optimal range of parameter β, the ratio of the damping coefficient of the additional damping support to the equivalent linear damping coefficient of isolation layer, to get a positive response of DISO system. The seismic response of the DISO structural system was analyzed and the correlation that the parameters including damping arrangement coefficient, damping index and damping coefficient have with the seismic response of the structure were discussed. The results show that the acceleration response of the DISO system decreases to minimum value with the increasing of damping coefficient, along with whiplash effect, and then increases regularly. There is a positive correlation between the damping coefficient and displacement decline rate. A case study has been completed. The numerical results show that the DISO system performs more effectively than the ISO system (refer to the traditional isolated structural system). Given that the equipment and pipelines in NPPs are sensitive to large isolation displacement, the displacement response of DISO system is less than that of the traditional isolated structural system without acceleration amplifying. Moreover, both acceleration of superstructure and displacement get decreased. And IDA-based (Incremental Dynamic Analysis) seismic fragility analysis of ISO and DISO system has been performed. DISO system has the smaller probability of failure compared with the ISO system.
{"title":"Seismic Behavior Study of a new Isolated Structure System With Damping Supporter for Nuclear Power Plants in a Cave or Underground","authors":"Yixin Feng, Haoqing Xu, Wenguang Liu, Qiang Zhang","doi":"10.1115/icone29-92634","DOIUrl":"https://doi.org/10.1115/icone29-92634","url":null,"abstract":"\u0000 There have been heated discussions on how to reduce the earthquake response of NPPs. Generally, isolation technology is one of the most efficient methods but researches on traditional isolated structure and 3D isolated structure for NPPs, to some extent, have deficiencies. Therefore, in this paper, a new structural system which included base isolation system and lateral damping supporters for NPPs (hereinafter referred to as DISO) in a cave or underground was proposed to obtain a positive seismic response, aiming to protect the nuclear island. The motion characteristics of the system have been studied and a simplified mathematical model has been set up. The dynamic equation was deduced and solved. The theoretical analysis results that there is an optimal range of parameter β, the ratio of the damping coefficient of the additional damping support to the equivalent linear damping coefficient of isolation layer, to get a positive response of DISO system. The seismic response of the DISO structural system was analyzed and the correlation that the parameters including damping arrangement coefficient, damping index and damping coefficient have with the seismic response of the structure were discussed. The results show that the acceleration response of the DISO system decreases to minimum value with the increasing of damping coefficient, along with whiplash effect, and then increases regularly. There is a positive correlation between the damping coefficient and displacement decline rate. A case study has been completed. The numerical results show that the DISO system performs more effectively than the ISO system (refer to the traditional isolated structural system). Given that the equipment and pipelines in NPPs are sensitive to large isolation displacement, the displacement response of DISO system is less than that of the traditional isolated structural system without acceleration amplifying. Moreover, both acceleration of superstructure and displacement get decreased. And IDA-based (Incremental Dynamic Analysis) seismic fragility analysis of ISO and DISO system has been performed. DISO system has the smaller probability of failure compared with the ISO system.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"39 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124370169","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}
Risk informed safety margin characteristics (RISMC) analysis Method, is a safety analysis method which combines probability safety analysis and determinism safety analysis of nuclear power plant. General RISMC analysis platform “CARS” (Coupling Platform for Advanced Risk Simulation) can provide parameter distribution and Monte Carlo sampling function, and interactive coupling function with data of the thermal hydraulic safety analysis system program. The paper uses CARS and thermal hydraulic safety analysis system program to analyzes the technical requirements of personnel training and engineering modification of nuclear power plant, by means of accident scenario sampling, physical mechanism calculation and scenario consequence statistics. In the analysis process, various important factors which may affect the event process and consequences are comprehensively considered through the analysis of nuclear power plant operation rules, equipment reliability and the uncertainty distribution characteristics of key factors. And consequences of event scenario development, nuclear power plant response and the time course are analyzed through establish a realistic physical model by using thermal hydraulic safety analysis system program. The calculation results give the target level of personnel training and the effectiveness of engineering modification. The results show that RISMC analysis Method is an effective safty analysis and decision making method for operation and maintenance of nuclear power plants.
{"title":"RISMC Analysis Method Used in Operation and Maintenance of Nuclear Power Plant","authors":"Zhao Wang, Weidong Liu, Leilei Liu, Yongxiang Wang, Liangwang Xu, Liang Pei, Qiongzhe Li, Jianbing Guo","doi":"10.1115/icone29-91667","DOIUrl":"https://doi.org/10.1115/icone29-91667","url":null,"abstract":"\u0000 Risk informed safety margin characteristics (RISMC) analysis Method, is a safety analysis method which combines probability safety analysis and determinism safety analysis of nuclear power plant. General RISMC analysis platform “CARS” (Coupling Platform for Advanced Risk Simulation) can provide parameter distribution and Monte Carlo sampling function, and interactive coupling function with data of the thermal hydraulic safety analysis system program. The paper uses CARS and thermal hydraulic safety analysis system program to analyzes the technical requirements of personnel training and engineering modification of nuclear power plant, by means of accident scenario sampling, physical mechanism calculation and scenario consequence statistics. In the analysis process, various important factors which may affect the event process and consequences are comprehensively considered through the analysis of nuclear power plant operation rules, equipment reliability and the uncertainty distribution characteristics of key factors. And consequences of event scenario development, nuclear power plant response and the time course are analyzed through establish a realistic physical model by using thermal hydraulic safety analysis system program. The calculation results give the target level of personnel training and the effectiveness of engineering modification. The results show that RISMC analysis Method is an effective safty analysis and decision making method for operation and maintenance of nuclear power plants.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123055506","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}
Ping Yang, Xianjia Huang, Lan Peng, Chaoliang Xing, Chunyang Zhao
Cables are one of the main fire hazards in nuclear power plants (NPPs). Cable arrangement, especially cable spacing in an electrical cable tray has a significant effect on cable tray fire. Predicting cable fire behavior in NPPs is necessary for the nuclear safety analysis. In this work, cable fire experiments with halogen-free flame retardant cable (XLPE cable with a diameter of 8.55mm) were carried out in a cone calorimeter. Cable spacing was changed from 0 mm to triple of cable diameter, including 7 different spacing arrangements under 3 different external heat fluxes. Besides, the experimental data of two other halogen-free flame retardant cables in the literature available are used. With the increase of cable spacing, the heat release rate per unit area (HRRPUA) was significantly increased and the fire duration was considerably decreased, while the time to ignition (TTI) of the cable was basically invariable with different cable spacings. Furthermore, the effects of the cable spacing on the cable burning become constant when the cable spacing was larger than a critical cable spacing. This work provides experimental support for industrial cable fire risk assessment, and the above findings are beneficial for fire prevention in the design stage of NPPs.
{"title":"Experimental Investigation of Cable Arrangement on the Cable Burning Behavior in a Cone Calorimeter","authors":"Ping Yang, Xianjia Huang, Lan Peng, Chaoliang Xing, Chunyang Zhao","doi":"10.1115/icone29-92145","DOIUrl":"https://doi.org/10.1115/icone29-92145","url":null,"abstract":"\u0000 Cables are one of the main fire hazards in nuclear power plants (NPPs). Cable arrangement, especially cable spacing in an electrical cable tray has a significant effect on cable tray fire. Predicting cable fire behavior in NPPs is necessary for the nuclear safety analysis. In this work, cable fire experiments with halogen-free flame retardant cable (XLPE cable with a diameter of 8.55mm) were carried out in a cone calorimeter. Cable spacing was changed from 0 mm to triple of cable diameter, including 7 different spacing arrangements under 3 different external heat fluxes. Besides, the experimental data of two other halogen-free flame retardant cables in the literature available are used. With the increase of cable spacing, the heat release rate per unit area (HRRPUA) was significantly increased and the fire duration was considerably decreased, while the time to ignition (TTI) of the cable was basically invariable with different cable spacings. Furthermore, the effects of the cable spacing on the cable burning become constant when the cable spacing was larger than a critical cable spacing. This work provides experimental support for industrial cable fire risk assessment, and the above findings are beneficial for fire prevention in the design stage of NPPs.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"72 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132483989","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}
High-temperature gas-cooled reactor (HTGR) is paid more attention in terms of inherent safety, high efficiency, and less liquid radioactive waste. The research on the source term of wastewater and the discharge methods of liquid effluent from HTGR can not only reduce the possible radiation risk to the environment and the public, but also promote the construction of inland nuclear power plants (NPPs), and improve the public acceptance of inland nuclear facilities. In this paper, calculation of activated corrosion products (ACPs) in the cooling water systems have been done in HTR-PM (High-Temperature Gas-Cooled Reactor Pebble-Bed Module) under normal operation conditions. Besides, the discharge methods of liquid effluent from inland NPPs without receiving water or with poor dilution conditions are studied. The calculation results indicate the cooling water of the heat removal system and shielded and component cooling water system needs to be changed once during the 40-year lifespan. The cooling water of the reactor support cooling system does not need to be changed. The annual effective dose resulting from liquid effluent through liquid-to-gas discharge using the data from a typical inland NPP site in China is less than direct discharge into the receiving water. The liquid-to-gas discharge is suitable for inland NPPs without receiving water or with poor dilution conditions.
{"title":"Research on Wastewater Source Term in High-Temperature Gas-Cooled Reactor and the Environmental Impact at Inland Sites","authors":"Yichun Zhang, Chuan Li, Haiyang Li, Jian-zhu Cao","doi":"10.1115/icone29-91467","DOIUrl":"https://doi.org/10.1115/icone29-91467","url":null,"abstract":"\u0000 High-temperature gas-cooled reactor (HTGR) is paid more attention in terms of inherent safety, high efficiency, and less liquid radioactive waste. The research on the source term of wastewater and the discharge methods of liquid effluent from HTGR can not only reduce the possible radiation risk to the environment and the public, but also promote the construction of inland nuclear power plants (NPPs), and improve the public acceptance of inland nuclear facilities. In this paper, calculation of activated corrosion products (ACPs) in the cooling water systems have been done in HTR-PM (High-Temperature Gas-Cooled Reactor Pebble-Bed Module) under normal operation conditions. Besides, the discharge methods of liquid effluent from inland NPPs without receiving water or with poor dilution conditions are studied. The calculation results indicate the cooling water of the heat removal system and shielded and component cooling water system needs to be changed once during the 40-year lifespan. The cooling water of the reactor support cooling system does not need to be changed. The annual effective dose resulting from liquid effluent through liquid-to-gas discharge using the data from a typical inland NPP site in China is less than direct discharge into the receiving water. The liquid-to-gas discharge is suitable for inland NPPs without receiving water or with poor dilution conditions.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"8584 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128707607","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}
Gao Tianyu, Zhou Tao, Liu Wenbin, Tang Jianyu, Lu Huaichang
Heat pipe cooled reactors have a promising future in the next generation of reactor design. The heat pipe, which is the core device for heat transfer from the core to the outside, determines the safety of the heat pipe reactor and the efficiency of power generation. In the high-temperature, high-pressure, high-irradiation environment of a nuclear reactor, the non-condensable gas produced by the lithium heat pipe - helium - affects the heat transfer efficiency of the heat pipe. Therefore, by modeling the heat transfer process of the heat pipe, the effect of the content of the non-condensable gas on the heat transfer performance of the heat pipe is calculated. It can be concluded that: at a certain input power, the temperature of both the hot end and the cold end of the heat pipe increases with the increase of the content of non-condensable gas; the distribution of non-condensable gas in the heat pipe is irregular, more at the hot end and less at the cold end; at a certain input power, the equivalent thermal resistance of the heat pipe increases with the increase of the content of non-condensable gas, and the higher the content of non-condensable gas, the greater the effect on the heat transfer effect of the heat pipe; in a certain range The effect of power on the equivalent thermal resistance of heat pipe is almost none.
{"title":"U Influence of Non-Condensable Gas in Mobile Heat Pipe Small Reactor on Heat Pipe Heat Transfer Performance","authors":"Gao Tianyu, Zhou Tao, Liu Wenbin, Tang Jianyu, Lu Huaichang","doi":"10.1115/icone29-89437","DOIUrl":"https://doi.org/10.1115/icone29-89437","url":null,"abstract":"\u0000 Heat pipe cooled reactors have a promising future in the next generation of reactor design. The heat pipe, which is the core device for heat transfer from the core to the outside, determines the safety of the heat pipe reactor and the efficiency of power generation. In the high-temperature, high-pressure, high-irradiation environment of a nuclear reactor, the non-condensable gas produced by the lithium heat pipe - helium - affects the heat transfer efficiency of the heat pipe. Therefore, by modeling the heat transfer process of the heat pipe, the effect of the content of the non-condensable gas on the heat transfer performance of the heat pipe is calculated. It can be concluded that: at a certain input power, the temperature of both the hot end and the cold end of the heat pipe increases with the increase of the content of non-condensable gas; the distribution of non-condensable gas in the heat pipe is irregular, more at the hot end and less at the cold end; at a certain input power, the equivalent thermal resistance of the heat pipe increases with the increase of the content of non-condensable gas, and the higher the content of non-condensable gas, the greater the effect on the heat transfer effect of the heat pipe; in a certain range The effect of power on the equivalent thermal resistance of heat pipe is almost none.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"44 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117022561","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}
Design basis earthquake may cause various internal hazards in nuclear power plants, such as fire, explosion, flooding, missiles, high-energy pipelines break, etc. In addition, earthquakes may also cause non-seismic items to fall or collapse, thus affecting safety-related systems or equipment. As the subsequent situation induced by earthquake is quite complex, no practicable methods have been presented yet to be used for the safety evaluation after earthquake. This paper introduces a feasible safety assessment concept for the above-mentioned hazards caused by the design basis earthquake based on the engineering practice experience. This general assessment concept based on the deterministic approach could be used for guidance of safety evaluation during or after earthquake for nuclear power plants.
{"title":"General Safety Assessment of Seismically Induced Hazards for NPPs","authors":"Guanghua Yang, Cuilian Yang","doi":"10.1115/icone29-93121","DOIUrl":"https://doi.org/10.1115/icone29-93121","url":null,"abstract":"\u0000 Design basis earthquake may cause various internal hazards in nuclear power plants, such as fire, explosion, flooding, missiles, high-energy pipelines break, etc. In addition, earthquakes may also cause non-seismic items to fall or collapse, thus affecting safety-related systems or equipment. As the subsequent situation induced by earthquake is quite complex, no practicable methods have been presented yet to be used for the safety evaluation after earthquake. This paper introduces a feasible safety assessment concept for the above-mentioned hazards caused by the design basis earthquake based on the engineering practice experience.\u0000 This general assessment concept based on the deterministic approach could be used for guidance of safety evaluation during or after earthquake for nuclear power plants.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128871883","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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