As the final heat sink of the nuclear power plant, the cool source is responsible for the safety function of the waste heat export of the reactor. The explosive invasion of marine organisms and other wastes have a great impact on the cold source system of nuclear power plants, which will lead to transient events, load reduction and shutdown of nuclear power unit, and even affect the availability of the cold source system in severe cases. which are huge threats and challenges for the safe and stable operation of unit. thereby reducing the power generation efficiency, safety, economy, and sustainable development of nuclear power plants. Clogging of the CFI drum net is one of the important factors causing cold source system accidents. this paper puts forward the solution idea of maintaining the water level of the forebay through the variable frequency pump, so as to ensure that the CRF pump keeps running state under the condition of cold source blockage, so as to maintain the function of the cold source system. and based on the full range simulator of nuclear power units, consider CFI drum network differential pressure high alarm failure combined with a monitoring device for the blockage degree of the drum net, the scheme is simulated and verified, and the load that the nuclear power unit can operate safely under different blocking degrees of drum network is given. the research results of this paper provide a solution for alleviating the cold source problem, and the solution can also provide a reference for the safety improvement of the cold source of actual unit and the design of new reactor.
{"title":"Simulation Research on Safe Load-Reducing Operation Margin of Nuclear Power Unit With Clogging of the CFI Drum Net","authors":"Jinchao Liu, Yaozu Lin, Bijun Chen, Xue Lyu, Jiayi Li, Heng Zhang","doi":"10.1115/icone29-92540","DOIUrl":"https://doi.org/10.1115/icone29-92540","url":null,"abstract":"\u0000 As the final heat sink of the nuclear power plant, the cool source is responsible for the safety function of the waste heat export of the reactor. The explosive invasion of marine organisms and other wastes have a great impact on the cold source system of nuclear power plants, which will lead to transient events, load reduction and shutdown of nuclear power unit, and even affect the availability of the cold source system in severe cases. which are huge threats and challenges for the safe and stable operation of unit. thereby reducing the power generation efficiency, safety, economy, and sustainable development of nuclear power plants. Clogging of the CFI drum net is one of the important factors causing cold source system accidents. this paper puts forward the solution idea of maintaining the water level of the forebay through the variable frequency pump, so as to ensure that the CRF pump keeps running state under the condition of cold source blockage, so as to maintain the function of the cold source system. and based on the full range simulator of nuclear power units, consider CFI drum network differential pressure high alarm failure combined with a monitoring device for the blockage degree of the drum net, the scheme is simulated and verified, and the load that the nuclear power unit can operate safely under different blocking degrees of drum network is given. the research results of this paper provide a solution for alleviating the cold source problem, and the solution can also provide a reference for the safety improvement of the cold source of actual unit and the design of new reactor.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"24 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":"114912735","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}
Huaichang Lu, T. Zhou, Wenbin Liu, Jianyu Tang, Tianyu Gao
Mobile heat pipe small reactors, with their simple structure, high mobility and inherent safety features, is a hot frontier topic for research at present. Based on Fluent software, numerical simulations are used to study the changes in the steady-state operating parameters of a high-temperature potassium heat pipe, and to investigate the effects of the horizontal acceleration, inclination angle and saturation temperature of the mass inside the heat pipe on the overall heat transfer performance of the heat pipe. It is found that the heat transfer performance of the heat pipe increases as the saturation temperature of the mass inside the tube decreases; the overall thermal resistance of the heat pipe tends to decrease and then increase slightly with increasing horizontal acceleration; the overall thermal resistance of the heat pipe tends to decrease and then increase with increasing tilt angle.
{"title":"Research on Heat Transfer Performance of Heat Pipe in Mobile Small Reactor","authors":"Huaichang Lu, T. Zhou, Wenbin Liu, Jianyu Tang, Tianyu Gao","doi":"10.1115/icone29-91655","DOIUrl":"https://doi.org/10.1115/icone29-91655","url":null,"abstract":"\u0000 Mobile heat pipe small reactors, with their simple structure, high mobility and inherent safety features, is a hot frontier topic for research at present. Based on Fluent software, numerical simulations are used to study the changes in the steady-state operating parameters of a high-temperature potassium heat pipe, and to investigate the effects of the horizontal acceleration, inclination angle and saturation temperature of the mass inside the heat pipe on the overall heat transfer performance of the heat pipe. It is found that the heat transfer performance of the heat pipe increases as the saturation temperature of the mass inside the tube decreases; the overall thermal resistance of the heat pipe tends to decrease and then increase slightly with increasing horizontal acceleration; the overall thermal resistance of the heat pipe tends to decrease and then increase with increasing tilt angle.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"5 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":"124137955","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Advanced Gas-Cooled Reactor (AGR) is the most common design of nuclear reactor in the UK. These reactors were commissioned in the 1980s and have now reached the end of their design life. As a type of thermal reactor, it uses graphite as a moderator to control the reaction. The cracking of graphite blocks in the core is an important factor limiting the lifetime of AGR stations. The damage of the AGR graphite will affect the power of the reactor and also become a hidden danger threatening the safety of the reactor. As a thermal reactor, it uses graphite as a moderator. Since graphite bricks cannot be replaced or repaired, in order to ensure the continued operation of the reactor, it is necessary to obtain timely and accurate damage information of graphite bricks. In order to carry out the research on in-situ multi-scale damage detection of AGR graphite, a linear CT imaging detection scheme was designed in combination with the core geometries of the AGR stations. Based on the GEANT4 Monte Carlo simulation software, an experiment is carried out to simulate a series of processes from signal acquisition to image reconstruction. Through the simulation experiment, the detection ability of the linear CT detection system under the interference of strong background radiation field signal in the AGR reactor and the geometric error of the detection system is verified, so as to evaluate the feasibility of the linear CT imaging detection scheme and guide the design of the detection system.
{"title":"Monte Carlo Simulation of In-Situ Multi-Scale Damage Detection of AGR Graphite","authors":"Tianchen Zeng, Yuewen Sun, Peng Cong","doi":"10.1115/icone29-92295","DOIUrl":"https://doi.org/10.1115/icone29-92295","url":null,"abstract":"\u0000 The Advanced Gas-Cooled Reactor (AGR) is the most common design of nuclear reactor in the UK. These reactors were commissioned in the 1980s and have now reached the end of their design life. As a type of thermal reactor, it uses graphite as a moderator to control the reaction. The cracking of graphite blocks in the core is an important factor limiting the lifetime of AGR stations. The damage of the AGR graphite will affect the power of the reactor and also become a hidden danger threatening the safety of the reactor. As a thermal reactor, it uses graphite as a moderator. Since graphite bricks cannot be replaced or repaired, in order to ensure the continued operation of the reactor, it is necessary to obtain timely and accurate damage information of graphite bricks.\u0000 In order to carry out the research on in-situ multi-scale damage detection of AGR graphite, a linear CT imaging detection scheme was designed in combination with the core geometries of the AGR stations. Based on the GEANT4 Monte Carlo simulation software, an experiment is carried out to simulate a series of processes from signal acquisition to image reconstruction. Through the simulation experiment, the detection ability of the linear CT detection system under the interference of strong background radiation field signal in the AGR reactor and the geometric error of the detection system is verified, so as to evaluate the feasibility of the linear CT imaging detection scheme and guide the design of the detection system.","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":"133249811","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}
We present the structure and characteristics of a photon-counting system for increased Steel Penetration (SP) ability in large container inspection area. Ideally, radiation images should count only unreacted photons. Conventional detectors utilize analog current integration and thus suffer from scattered signals and electronic noise which can deteriorate image quality. However, with the increasing demand for container inspection, higher requirements have been placed on image quality, scan speed as well as radiation dose. Therefore, we propose a new photon-counting mode large container inspection system that counts individual signals, in which scattered signals and electronic noise can be eliminated by setting an energy threshold. In this paper, we give a system overview and describe the key technologies in system design. Then the improvement of the presented system is analyzed theoretically from the aspect of physical interactions that take place in the system. Simulation of the Steel Penetration model is conducted in Geant4 software. Results show that the improvement of SP ability compared with the conventional current-integration system is about 3 to 4cm Fe even if we increase the scan speed or reduce the radiation intensity, and can be 1 to 2cm more by setting an appropriate energy threshold. Along with this property, the possibility of reducing radiation dose and increasing scan speed makes this system unique for future growth of container inspection missions.
{"title":"Improvement of Large Container Inspection System in Photon- Counting Mode Proved With Steel Penetration Ability","authors":"Huaxia Zhang, Zhifang Wu, Tao Dong, Shibo Jiang","doi":"10.1115/icone29-88874","DOIUrl":"https://doi.org/10.1115/icone29-88874","url":null,"abstract":"\u0000 We present the structure and characteristics of a photon-counting system for increased Steel Penetration (SP) ability in large container inspection area. Ideally, radiation images should count only unreacted photons. Conventional detectors utilize analog current integration and thus suffer from scattered signals and electronic noise which can deteriorate image quality. However, with the increasing demand for container inspection, higher requirements have been placed on image quality, scan speed as well as radiation dose. Therefore, we propose a new photon-counting mode large container inspection system that counts individual signals, in which scattered signals and electronic noise can be eliminated by setting an energy threshold.\u0000 In this paper, we give a system overview and describe the key technologies in system design. Then the improvement of the presented system is analyzed theoretically from the aspect of physical interactions that take place in the system. Simulation of the Steel Penetration model is conducted in Geant4 software. Results show that the improvement of SP ability compared with the conventional current-integration system is about 3 to 4cm Fe even if we increase the scan speed or reduce the radiation intensity, and can be 1 to 2cm more by setting an appropriate energy threshold. Along with this property, the possibility of reducing radiation dose and increasing scan speed makes this system unique for future growth of container inspection missions.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"54 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":"134212409","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The safety of I&C system of nuclear facilities is one of the key points of nuclear safety supervision. Due to the use of proprietary technology and generally not connected with IT systems, traditional I&C system of nuclear facilities mainly focuses on functional safety in design. Relevant standards such as IEC 61508, IEC 61511 and IEC 61513 all focus on ensuring functional safety. The extensive use of information technology in nuclear energy field has brought about cyber security threats, and cyber-attacks against nuclear facilities have been reported occasionally. The International Atomic Energy Agency (IAEA) and nuclear safety regulators around the world have issued laws, regulations and technical standards to strengthen the supervision of cyber security of nuclear facilities. However, how to properly deal with the relationship between functional safety and cyber security has become a hot issue when strengthening the supervision of cyber security of nuclear facilities. Based on the relevant international standards and national practices, this paper further clarifies the relevant concepts, review current construction practices, analyzes and discusses the relationship between functional safety and cyber security. The research lays a foundation for the subsequent establishment of applicable nuclear power cyber security standard system.
{"title":"Discussion on Functional Safety and Cyber Security of I&C System in Nuclear Facilities","authors":"Anyi Yang","doi":"10.1115/icone29-90807","DOIUrl":"https://doi.org/10.1115/icone29-90807","url":null,"abstract":"\u0000 The safety of I&C system of nuclear facilities is one of the key points of nuclear safety supervision. Due to the use of proprietary technology and generally not connected with IT systems, traditional I&C system of nuclear facilities mainly focuses on functional safety in design. Relevant standards such as IEC 61508, IEC 61511 and IEC 61513 all focus on ensuring functional safety. The extensive use of information technology in nuclear energy field has brought about cyber security threats, and cyber-attacks against nuclear facilities have been reported occasionally. The International Atomic Energy Agency (IAEA) and nuclear safety regulators around the world have issued laws, regulations and technical standards to strengthen the supervision of cyber security of nuclear facilities. However, how to properly deal with the relationship between functional safety and cyber security has become a hot issue when strengthening the supervision of cyber security of nuclear facilities. Based on the relevant international standards and national practices, this paper further clarifies the relevant concepts, review current construction practices, analyzes and discusses the relationship between functional safety and cyber security. The research lays a foundation for the subsequent establishment of applicable nuclear power cyber security standard system.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"4 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":"129312318","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
With the development of nuclear energy and the need for environmental protection, some advanced small reactors have gradually entered the public’s field of vision. The passive natural circulation is an important part of the inherent safety feature for advanced small reactors. Based on the limitations of the site environment, the large heat exchange area of the air cooler, and the high cost of equipment construction for the air-cooled waste heat removal scheme of a small reactor, this paper studies the water-cooled passive residual heat removal scheme and analyzes the water-cooled water cooling by establishing a system natural circulation hydraulic calculation model. From the aspects of equipment scale, construction cost, and operation & maintenance cost, etc., the advantages and disadvantages of several schemes were analyzed. This study shows that the water-cooled passive waste heat removal scheme can extract the core heat under accident conditions; it can reduce the equipment volume, and reduce the cost. Finally, suggestions are put forward for the follow-up improvement direction, and further research and realization of steam-water two-phase condensation heat transfer are made to improve heat exchange efficiency, reduce equipment scale, and improve the economy, which can provide a reference for the follow-up small reactor research and development.
{"title":"Feasibility Study of Passive Water-Cooled Residual Heat Removal Scheme for Small Reactors","authors":"Fei Yu, Yueming Fu, Feng Zhu, Mingqian Zhang","doi":"10.1115/icone29-92779","DOIUrl":"https://doi.org/10.1115/icone29-92779","url":null,"abstract":"\u0000 With the development of nuclear energy and the need for environmental protection, some advanced small reactors have gradually entered the public’s field of vision. The passive natural circulation is an important part of the inherent safety feature for advanced small reactors. Based on the limitations of the site environment, the large heat exchange area of the air cooler, and the high cost of equipment construction for the air-cooled waste heat removal scheme of a small reactor, this paper studies the water-cooled passive residual heat removal scheme and analyzes the water-cooled water cooling by establishing a system natural circulation hydraulic calculation model. From the aspects of equipment scale, construction cost, and operation & maintenance cost, etc., the advantages and disadvantages of several schemes were analyzed.\u0000 This study shows that the water-cooled passive waste heat removal scheme can extract the core heat under accident conditions; it can reduce the equipment volume, and reduce the cost. Finally, suggestions are put forward for the follow-up improvement direction, and further research and realization of steam-water two-phase condensation heat transfer are made to improve heat exchange efficiency, reduce equipment scale, and improve the economy, which can provide a reference for the follow-up small reactor research and development.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"51 2 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":"127396006","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}
Mengyan Song, Di Jiang, Changkai Li, C. Zhang, X. Zhi
Since the 9.11 incident, the impact of large commercial airliners on nuclear power plants and other important national facilities has attracted worldwide attention. The safety of flying objects hitting nuclear containment has become one of the important research topics in the field of nuclear power generation. In this paper, the impact resistance of non-prestressed and prestressed concrete slabs is studied through detailed experiments and numerical studies. By non-bonding rear sheeting, prestress of about 10MPa is applied to concrete slabs measuring 1400mm × 600mm, which are then hit by hemispherical steel hammers (590kg) at heights of 2m and 4m above the center of the span. The impact force, displacement, and the reinforced concrete slab and the prestressed concrete slab are compared. In LS-DYNA finite element software, the CSCM concrete structure model is used to study the numerical values of the experiments carried out. It has been found that with the increase of impact energy, the deformation produced by reinforced concrete slabs increases and is not recoverable. On the other hand, after applying presession, the impact force of prestressed concrete slabs was increased by 53.8% and displacement was reduced by 45% due to the increase of the stiffness of the plate. The maximum reproduction deviation of the impact force and displacement values is 17% and 14%, respectively. The research results will be applied to the structural dynamic impact protection design of nuclear power plant containment.
{"title":"Study on the Response of Prestressed Reinforced Concrete Slabs With High Reinforcement Ratio Under Low Speed Impact","authors":"Mengyan Song, Di Jiang, Changkai Li, C. Zhang, X. Zhi","doi":"10.1115/icone29-92808","DOIUrl":"https://doi.org/10.1115/icone29-92808","url":null,"abstract":"\u0000 Since the 9.11 incident, the impact of large commercial airliners on nuclear power plants and other important national facilities has attracted worldwide attention. The safety of flying objects hitting nuclear containment has become one of the important research topics in the field of nuclear power generation. In this paper, the impact resistance of non-prestressed and prestressed concrete slabs is studied through detailed experiments and numerical studies. By non-bonding rear sheeting, prestress of about 10MPa is applied to concrete slabs measuring 1400mm × 600mm, which are then hit by hemispherical steel hammers (590kg) at heights of 2m and 4m above the center of the span. The impact force, displacement, and the reinforced concrete slab and the prestressed concrete slab are compared. In LS-DYNA finite element software, the CSCM concrete structure model is used to study the numerical values of the experiments carried out. It has been found that with the increase of impact energy, the deformation produced by reinforced concrete slabs increases and is not recoverable. On the other hand, after applying presession, the impact force of prestressed concrete slabs was increased by 53.8% and displacement was reduced by 45% due to the increase of the stiffness of the plate. The maximum reproduction deviation of the impact force and displacement values is 17% and 14%, respectively. The research results will be applied to the structural dynamic impact protection design of nuclear power plant containment.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"76 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":"115964729","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}
Dongyang Wang, Xiaoying Sun, Ziqiao Liu, Yingying Gan
The proposed site condition of an underground storage facility of nuclear power project (NPP) is complex, characterized by strong spatial variability of bedrock surface, non-uniform soil layers and underlying soft soil. If the substructure method commonly used in nuclear power engineering seismic analysis is adopted, the non-homogeneous foundation must be simplified as the horizontal layered site condition. Therefore, the dynamic interaction features among piles, foundation soil and superstructures under strong earthquakes cannot be reflected accurately. Taking it as the project background, this paper conducts a research on the three-dimensional time-history seismic analysis method for NPPs. On the theoretical basis of viscoelastic artificial boundary and wave equation method, the soil-structure dynamic interaction analysis methodology in complicated non-homogeneous site is put forward in this paper, which is called the integral time-domain method. And an efficient platform is set up for computing artificial boundary parameters and seismic motions. Then the seismic response analysis of pile foundation of the underground storage facility is completed. It is concluded that the pile-soil-structure interaction does not affect horizontal response of the super structure significantly but leads to a frequency shift of the response peak in vertical direction. Additionally, the substructure method is used to verify the accuracy of the proposed method and the results of two methods match well. In conclusion, the integral time-domain method contributes to appropriately simulating the foundation radiation damping and seismic input. It has high accuracy and applicability. The research method and the computing platform can be taken as technical references for seismic response analysis of NPPs constructed in non-horizontal layered foundation or complex geologic bodies conditions.
{"title":"Seismic Response Analysis of Pile Foundation of an Underground Storage Facility of Nuclear Power Project","authors":"Dongyang Wang, Xiaoying Sun, Ziqiao Liu, Yingying Gan","doi":"10.1115/icone29-89349","DOIUrl":"https://doi.org/10.1115/icone29-89349","url":null,"abstract":"\u0000 The proposed site condition of an underground storage facility of nuclear power project (NPP) is complex, characterized by strong spatial variability of bedrock surface, non-uniform soil layers and underlying soft soil. If the substructure method commonly used in nuclear power engineering seismic analysis is adopted, the non-homogeneous foundation must be simplified as the horizontal layered site condition. Therefore, the dynamic interaction features among piles, foundation soil and superstructures under strong earthquakes cannot be reflected accurately. Taking it as the project background, this paper conducts a research on the three-dimensional time-history seismic analysis method for NPPs. On the theoretical basis of viscoelastic artificial boundary and wave equation method, the soil-structure dynamic interaction analysis methodology in complicated non-homogeneous site is put forward in this paper, which is called the integral time-domain method. And an efficient platform is set up for computing artificial boundary parameters and seismic motions. Then the seismic response analysis of pile foundation of the underground storage facility is completed. It is concluded that the pile-soil-structure interaction does not affect horizontal response of the super structure significantly but leads to a frequency shift of the response peak in vertical direction. Additionally, the substructure method is used to verify the accuracy of the proposed method and the results of two methods match well. In conclusion, the integral time-domain method contributes to appropriately simulating the foundation radiation damping and seismic input. It has high accuracy and applicability. The research method and the computing platform can be taken as technical references for seismic response analysis of NPPs constructed in non-horizontal layered foundation or complex geologic bodies conditions.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"403 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":"132033609","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}
Nuclear material license holder should develop a nuclear material accounting management system and achieve closed balance accountancy. According to regulations requirements, nuclear material accountancy implementation is based on physical inventory and material measurement. Generally, the spent fuel reprocessing plant operates uninterrupted, its annual throughput of nuclear materials is huge, there are different kinds of measurement and analysis techniques configured combing the reprocessing process. So the closed balance approach of nuclear materials accountancy is a great challenge for reprocessing plant holders. In order to improve the accuracy and reliability of nuclear material, enhance the timeliness of abnormal detection, near-real time accounting prospect of spent fuel reprocessing plant was researched. In this paper, the crucial aspects affecting closed balance accountancy was discussed, the factors such as the head-end receiver-shipper inventory difference, the on-line process monitoring accuracy, the applicability of the international target value of nuclear material measurement uncertainty, and the nuclear material balance model of reprocessing plants. As summarized, proposed suggestions and solutions for nuclear material balance in spent fuel reprocessing plant was put forward on the end of the paper.
{"title":"Implementation Challenges and Proposed Suggestions for Nuclear Material Accountancy Management in Spent Fuel Reprocessing Plant","authors":"Lixia He, Lei Bai, Qiang Miao, Qun Yang","doi":"10.1115/icone29-91491","DOIUrl":"https://doi.org/10.1115/icone29-91491","url":null,"abstract":"\u0000 Nuclear material license holder should develop a nuclear material accounting management system and achieve closed balance accountancy. According to regulations requirements, nuclear material accountancy implementation is based on physical inventory and material measurement. Generally, the spent fuel reprocessing plant operates uninterrupted, its annual throughput of nuclear materials is huge, there are different kinds of measurement and analysis techniques configured combing the reprocessing process. So the closed balance approach of nuclear materials accountancy is a great challenge for reprocessing plant holders. In order to improve the accuracy and reliability of nuclear material, enhance the timeliness of abnormal detection, near-real time accounting prospect of spent fuel reprocessing plant was researched. In this paper, the crucial aspects affecting closed balance accountancy was discussed, the factors such as the head-end receiver-shipper inventory difference, the on-line process monitoring accuracy, the applicability of the international target value of nuclear material measurement uncertainty, and the nuclear material balance model of reprocessing plants. As summarized, proposed suggestions and solutions for nuclear material balance in spent fuel reprocessing plant was put forward on the end of the paper.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"16 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":"132083159","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}
All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of a nuclear leakage accident, radionuclides will diffuse through the atmosphere and the ocean, and pollute the ecological environment. To solve the problem of radionuclide diffusion in the ocean, we first need to solve the source term problem. The study on source characteristics of radionuclide diffusion in the ocean under oceanic radioactive events plays an important role in simulation and emergency decision-making. The meaning of source term and unit parameters were given, and types, forms and release modes of radionuclides were analyzed under nuclear leakage accidents. The storage and release of radionuclides were calculated. Then the dry and wet deposition fluxes of radionuclides were calculated. The results show that 131I, 134Cs and 137Cs are the most important radionuclides in radiation dose evaluation and they will be in the form of CSI and CsOH under nuclear leakage accidents. Radionuclides are released into the environment in the form of micron aerosol. Radionuclides entering the ocean include direct release and atmospheric radionuclide deposition. The dry deposition flux is 3 orders to 4 orders of magnitude lower than the ground radionuclide concentration. The wet deposition flux within 300m is much larger than the dry deposition flux, and the wet deposition flux outside 300m is about 1 order to 2 orders of magnitude larger than the dry deposition flux. It provides a prerequisite for the calculation of radionuclide diffusion in the ocean under oceanic radioactive events.
{"title":"Research on Source Characteristics of Radionuclide Diffusion in the Ocean Under Oceanic Radioactive Events","authors":"Zichao Li, Rong-chang Chen, Chen Liu, Qingqing Xue, Zhixia Wang, Tao Zhou","doi":"10.1115/icone29-89248","DOIUrl":"https://doi.org/10.1115/icone29-89248","url":null,"abstract":"\u0000 All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of a nuclear leakage accident, radionuclides will diffuse through the atmosphere and the ocean, and pollute the ecological environment. To solve the problem of radionuclide diffusion in the ocean, we first need to solve the source term problem. The study on source characteristics of radionuclide diffusion in the ocean under oceanic radioactive events plays an important role in simulation and emergency decision-making. The meaning of source term and unit parameters were given, and types, forms and release modes of radionuclides were analyzed under nuclear leakage accidents. The storage and release of radionuclides were calculated. Then the dry and wet deposition fluxes of radionuclides were calculated. The results show that 131I, 134Cs and 137Cs are the most important radionuclides in radiation dose evaluation and they will be in the form of CSI and CsOH under nuclear leakage accidents. Radionuclides are released into the environment in the form of micron aerosol. Radionuclides entering the ocean include direct release and atmospheric radionuclide deposition. The dry deposition flux is 3 orders to 4 orders of magnitude lower than the ground radionuclide concentration. The wet deposition flux within 300m is much larger than the dry deposition flux, and the wet deposition flux outside 300m is about 1 order to 2 orders of magnitude larger than the dry deposition flux. It provides a prerequisite for the calculation of radionuclide diffusion in the ocean under oceanic radioactive events.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"137 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":"131749332","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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