Zhenying Jiao, Zhen-Ying Wang, Yu Huang, Liu Liu, Taike Wang
The post-accident operation status of reactor coolant pumps (RCP) play an important rule in accident mitigation, especially for the loss of coolant accident (LOCA), the prolonged RCP operation after medium and small LOCA will aggravate the deterioration process of RCS water inventory. This paper first analyzes the adverse influence of delayed trip of RCPs after medium and small break on RCS, then the principle of determining the trip criteria of RCPs is suggested, the trip criteria of RCPs from the worldwide nuclear steam supply system venders are investigated and summarized. From the perspective of prolonging the operator ‘s non-intervention time, reasonable measures to alleviate the consequences of medium and small LOCA, such as the RCP automatic trip logic and RCS fast depressurization by cooling, etc., are of great significance to improving the safety level of the plant.
{"title":"Review on Reactor Coolant Pumps Trip Criteria After Small Break LOCA in PWR","authors":"Zhenying Jiao, Zhen-Ying Wang, Yu Huang, Liu Liu, Taike Wang","doi":"10.1115/icone29-91395","DOIUrl":"https://doi.org/10.1115/icone29-91395","url":null,"abstract":"\u0000 The post-accident operation status of reactor coolant pumps (RCP) play an important rule in accident mitigation, especially for the loss of coolant accident (LOCA), the prolonged RCP operation after medium and small LOCA will aggravate the deterioration process of RCS water inventory. This paper first analyzes the adverse influence of delayed trip of RCPs after medium and small break on RCS, then the principle of determining the trip criteria of RCPs is suggested, the trip criteria of RCPs from the worldwide nuclear steam supply system venders are investigated and summarized. From the perspective of prolonging the operator ‘s non-intervention time, reasonable measures to alleviate the consequences of medium and small LOCA, such as the RCP automatic trip logic and RCS fast depressurization by cooling, etc., are of great significance to improving the safety level of the plant.","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":"128549146","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}
Lan Peng, Xianjia Huang, Jinkai Wang, He Zhu, Ping Yang, Chaoliang Xing, Chunyang Zhao
The heat release rate of the fire source is one of the most important parameters for the fire hazards analysis in nuclear power plants. For a fire in a confined compartment, the oxygen concentration has a critical effect on the fire heat release rate under oxygen-deficient situation. In the present work, a modified two-zone fire model was developed to predict the pool fire heat release rate in the oxygen-deficient situation. A simple model estimating the effect of the oxygen concentration on the fire heat release rate was incorporated into two-zone model, CFAST. Furthermore, the conservation of fuel mass was also taken into consideration. Experimental results from three fire experiments of pool fire under the oxygen-deficient situation in the literature available were used to validate the reliability of the modified two-zone model. The oxygen concentration in the compartment was decreased and became oxygen-deficient condition due to the mechanical ventilation. Compared with the original CFAST model, the accuracy of predicting pool fire under oxygen-deficient situation was increased considerably by the modified model, especially for the pool fire at steady under oxygen-deficient situation. In comparison with the experimental data, the cosine similarities of the three heat release rate curves were all over 0.95 and the max relative error of the burning time was 41.9%, which is significantly lower than that of the original two-zone model prediction.
{"title":"A Modified Two-Zone Model for Predicting Heat Release Rate of Pool Fire in a Confined Space","authors":"Lan Peng, Xianjia Huang, Jinkai Wang, He Zhu, Ping Yang, Chaoliang Xing, Chunyang Zhao","doi":"10.1115/icone29-92422","DOIUrl":"https://doi.org/10.1115/icone29-92422","url":null,"abstract":"\u0000 The heat release rate of the fire source is one of the most important parameters for the fire hazards analysis in nuclear power plants. For a fire in a confined compartment, the oxygen concentration has a critical effect on the fire heat release rate under oxygen-deficient situation. In the present work, a modified two-zone fire model was developed to predict the pool fire heat release rate in the oxygen-deficient situation. A simple model estimating the effect of the oxygen concentration on the fire heat release rate was incorporated into two-zone model, CFAST. Furthermore, the conservation of fuel mass was also taken into consideration. Experimental results from three fire experiments of pool fire under the oxygen-deficient situation in the literature available were used to validate the reliability of the modified two-zone model. The oxygen concentration in the compartment was decreased and became oxygen-deficient condition due to the mechanical ventilation. Compared with the original CFAST model, the accuracy of predicting pool fire under oxygen-deficient situation was increased considerably by the modified model, especially for the pool fire at steady under oxygen-deficient situation. In comparison with the experimental data, the cosine similarities of the three heat release rate curves were all over 0.95 and the max relative error of the burning time was 41.9%, which is significantly lower than that of the original two-zone model prediction.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"61 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":"125828709","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 ITER project, China is preparing to build China Fusion Engineering Experimental Reactor (CFETR), at present; there is no domestic nuclear fusion regulation framework. The nuclear safety regulation framework based on the fission technology is relatively solid and referable, which can provide reference for the establishment of fusion regulation framework. HAF102 “Safety Code on Nuclear Power Plant Design” is one of the most important regulations in Chinese nuclear safety framework. HAF102 specifies the nuclear safety requirements and provides guidance for the design of fission reactors in terms of defense in depth, safety functions and safety classification, these design principles are also crucial for fusion reactors. Therefore, this paper analyzes the general applicability of fusion reactors and HAF102-2016, identifies the non-conformance terms and categories the terms into generally applicable and partially applicable. The application analysis aims to identify the important safety issues that should be considered in the early stage of the development of fusion reactors, in addition, the inapplicable provisions of HAF102 are identified based on the characteristics of fusion reactors. The analysis lays a preliminary foundation for the establishment of a fusion regulation framework on safety issues in China, and on the other hand, it provides recommendations for the design principles of fusion reactors.
{"title":"The Applicability Analysis of Fusion Reactors and China Nuclear Safety Code HAF102","authors":"Yijie Qian, Chong Li","doi":"10.1115/icone29-91692","DOIUrl":"https://doi.org/10.1115/icone29-91692","url":null,"abstract":"\u0000 With the development of ITER project, China is preparing to build China Fusion Engineering Experimental Reactor (CFETR), at present; there is no domestic nuclear fusion regulation framework. The nuclear safety regulation framework based on the fission technology is relatively solid and referable, which can provide reference for the establishment of fusion regulation framework. HAF102 “Safety Code on Nuclear Power Plant Design” is one of the most important regulations in Chinese nuclear safety framework. HAF102 specifies the nuclear safety requirements and provides guidance for the design of fission reactors in terms of defense in depth, safety functions and safety classification, these design principles are also crucial for fusion reactors. Therefore, this paper analyzes the general applicability of fusion reactors and HAF102-2016, identifies the non-conformance terms and categories the terms into generally applicable and partially applicable. The application analysis aims to identify the important safety issues that should be considered in the early stage of the development of fusion reactors, in addition, the inapplicable provisions of HAF102 are identified based on the characteristics of fusion reactors. The analysis lays a preliminary foundation for the establishment of a fusion regulation framework on safety issues in China, and on the other hand, it provides recommendations for the design principles of fusion reactors.","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":"128935863","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}
Yu Fan, B. Jie, Xu Botao, Lv Xingbing, Li Yong, Zhang Xueqing
Marine biological outbreaks in coastal waters of coastal nuclear power plants have caused abnormal water intake systems of nuclear power plants for many times, and the risk Valuation method of marine biological outbreaks to nuclear power safety is still in the research and exploration stage. For nuclear power plants, the outbreak of marine organisms near the water intake is a natural event that affects the safety of nuclear power. According to the basic framework of “four-step method” on risk Valuation of National Academy of Sciences of the United States, aiming at the disaster risk caused by marine organisms that potentially affects the reliability of cold sources of nuclear power plants, the risk characterization method is studied, and the design protection measures are put forward.
{"title":"Research on Risk Characterization Method of Marine Biological Disaster Affecting Water Intake Safety of Nuclear Power Plant and Application of Design Protection","authors":"Yu Fan, B. Jie, Xu Botao, Lv Xingbing, Li Yong, Zhang Xueqing","doi":"10.1115/icone29-92742","DOIUrl":"https://doi.org/10.1115/icone29-92742","url":null,"abstract":"\u0000 Marine biological outbreaks in coastal waters of coastal nuclear power plants have caused abnormal water intake systems of nuclear power plants for many times, and the risk Valuation method of marine biological outbreaks to nuclear power safety is still in the research and exploration stage. For nuclear power plants, the outbreak of marine organisms near the water intake is a natural event that affects the safety of nuclear power. According to the basic framework of “four-step method” on risk Valuation of National Academy of Sciences of the United States, aiming at the disaster risk caused by marine organisms that potentially affects the reliability of cold sources of nuclear power plants, the risk characterization method is studied, and the design protection measures are put forward.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"25 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":"115110260","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 a typical type of the fourth-generation nuclear power system. The main supporting structure, consisting of graphite and carbon components, play a vital role in the construction of the HTGR. The quality of the components is essential for the safety operation of HTGR since they are irreplaceable throughout the reactor lifetime. The manufacture of the components is complex, including multiple process, during which defects such as holes and crack often arise inevitably. These defects may bring serious risk to the structural safety and steady operation of the reactor. Therefore, it is of great significance to inspect and evaluate the quality of the components. Considering the large size of the components as well as the long production cycle, traditional non-destructive testing method such as x-ray and eddy current testing are not applicable. Visual inspection and spot check are generally applied to check the surface condition, which are unable to provide the internal situation of the components. This paper proposes a helical CT based defects detection method for large size graphite and carbon components in HTGR. Graphite and carbon samples with artificial and original defect were produced, and various experiments were conducted on a multi-slice helical CT system to check the performance as well as optimize the operation parameter. The results indicates that defect larger than 2 mm in graphite components and 1 mm in carbon components can be detected and clearly visualized, which proves the feasibility of the proposed method.
{"title":"Defect Detection Method for Large Size Graphite and Carbon Components of High Temperature Gas Cooled Reactor Based on Computed Tomography","authors":"Renjie Liu, Yuewen Sun, Tianchen Zeng, Ximing Liu, Libin Sun, Li Shi, Peng Cong","doi":"10.1115/icone29-92345","DOIUrl":"https://doi.org/10.1115/icone29-92345","url":null,"abstract":"\u0000 High temperature gas cooled reactor (HTGR) is a typical type of the fourth-generation nuclear power system. The main supporting structure, consisting of graphite and carbon components, play a vital role in the construction of the HTGR. The quality of the components is essential for the safety operation of HTGR since they are irreplaceable throughout the reactor lifetime. The manufacture of the components is complex, including multiple process, during which defects such as holes and crack often arise inevitably. These defects may bring serious risk to the structural safety and steady operation of the reactor. Therefore, it is of great significance to inspect and evaluate the quality of the components.\u0000 Considering the large size of the components as well as the long production cycle, traditional non-destructive testing method such as x-ray and eddy current testing are not applicable. Visual inspection and spot check are generally applied to check the surface condition, which are unable to provide the internal situation of the components. This paper proposes a helical CT based defects detection method for large size graphite and carbon components in HTGR. Graphite and carbon samples with artificial and original defect were produced, and various experiments were conducted on a multi-slice helical CT system to check the performance as well as optimize the operation parameter. The results indicates that defect larger than 2 mm in graphite components and 1 mm in carbon components can be detected and clearly visualized, which proves the feasibility of the proposed method.","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":"132976365","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 modern automatic control technology, digital control technology has advanced technology and inherent advantages compared with traditional analog control technology. Therefore, the use of digital instrument and control system has become an inevitable trend of global nuclear power construction. Nuclear power embedded real-time operating system is an important part of digital instrument and control system in nuclear power plant. It is responsible for managing the hardware resources of the whole system and providing operation environment for application software. Its safety directly affects the safety of instrument and control system. This paper studies the application of embedded information technology in the safety of nuclear power plant, and puts forward a safety related scheme of embedded information system. The scheme fully considers the confidentiality, integrity and availability of information in the digital instrument and control system of nuclear power plant. The application results show that the scheme can meet the information security requirements of nuclear power instrument and control system, and has reference and guiding significance for the research and development of nuclear power embedded real-time operating system.
{"title":"Information Security Design of Nuclear Power Embedded Real-Time Operating System","authors":"Meng Li, G. Shi, Zhonggang Ma, Xiaowei Wang, Lichen Fan, Weiwei Dou","doi":"10.1115/icone29-89240","DOIUrl":"https://doi.org/10.1115/icone29-89240","url":null,"abstract":"\u0000 With the development of modern automatic control technology, digital control technology has advanced technology and inherent advantages compared with traditional analog control technology. Therefore, the use of digital instrument and control system has become an inevitable trend of global nuclear power construction. Nuclear power embedded real-time operating system is an important part of digital instrument and control system in nuclear power plant. It is responsible for managing the hardware resources of the whole system and providing operation environment for application software. Its safety directly affects the safety of instrument and control system. This paper studies the application of embedded information technology in the safety of nuclear power plant, and puts forward a safety related scheme of embedded information system. The scheme fully considers the confidentiality, integrity and availability of information in the digital instrument and control system of nuclear power plant. The application results show that the scheme can meet the information security requirements of nuclear power instrument and control system, and has reference and guiding significance for the research and development of nuclear power embedded real-time operating system.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"20 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":"133019944","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}
Yong Zhang, Zhao Zhang, Zi-Chun Wang, Yongtao Zhou, Bo-Wen Tang, Min-Hua Su, Wen-Bo Luo
To extend the intervals of Safety-related Instrument Calibration Periodic Test can shorten the refueling outage period and improve unit economy. Based on the general requirements for extending the Safety-related Instrument Calibration Periodic Test Intervals and in consideration of the characteristics of M310 Nuclear Power Unit with analog I&C System, an evaluation method is proposed and investigated, which consist of instrument drift analysis, protection channel analysis, channel uncertainty calculation, setpoint analysis and relevant channel tests evaluation. The research on the extension of steam generator level transmitter calibration test interval is carried out in detail. The results show that the drift resulting from the interval extension will has no effect on the protection channel setpoint, the acceptance criteria for channel consistency test and channel function test should remain unchanged. The research of this paper has a reference value for unit economy improvement and outage optimization.
{"title":"Study on the Method for the Safety-Related Instrument Calibration Surveillance Interval Extension of Nuclear Power Unit With Analog I&C System Based on Setpoint and Uncertainty Analysis","authors":"Yong Zhang, Zhao Zhang, Zi-Chun Wang, Yongtao Zhou, Bo-Wen Tang, Min-Hua Su, Wen-Bo Luo","doi":"10.1115/icone29-92941","DOIUrl":"https://doi.org/10.1115/icone29-92941","url":null,"abstract":"\u0000 To extend the intervals of Safety-related Instrument Calibration Periodic Test can shorten the refueling outage period and improve unit economy. Based on the general requirements for extending the Safety-related Instrument Calibration Periodic Test Intervals and in consideration of the characteristics of M310 Nuclear Power Unit with analog I&C System, an evaluation method is proposed and investigated, which consist of instrument drift analysis, protection channel analysis, channel uncertainty calculation, setpoint analysis and relevant channel tests evaluation. The research on the extension of steam generator level transmitter calibration test interval is carried out in detail. The results show that the drift resulting from the interval extension will has no effect on the protection channel setpoint, the acceptance criteria for channel consistency test and channel function test should remain unchanged. The research of this paper has a reference value for unit economy improvement and outage optimization.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"26 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":"128106864","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}
Xinyu Zhang, Bing-feng Dong, Hui Yang, Wentao Zhou, Dezhong Wang
10B has a high neutron absorption cross section, so boric acid is generally used as a neutron absorber for reactor control. During an emergency of a nuclear power system, boric acid water is injected into the reactor core by safety injection system to shut down the reactor. Boron concentration has an important impact on the operation safety of the power plant. The mixing process of boric acid and water is influenced by diffusion, convection and turbulence. A new laser-induced fluorescence method is proposed, which can accurately and quickly measure the concentration field of boric acid. This paper introduces the measurement method and the experimental study of boric acid diffusion in T-tube, and compares it with the measurement of wire mesh sensor. In addition, the CFD method is used to calculate the experimental conditions. The results obtained by the three methods were compared with each other, and the diffusion law and research of boric acid are obtained. This study can provide a certain basis for the follow-up study of boron diffusion.
{"title":"Simulation and Experimental Verification of Boron Diffusion and Mixing in T-Tube Flow","authors":"Xinyu Zhang, Bing-feng Dong, Hui Yang, Wentao Zhou, Dezhong Wang","doi":"10.1115/icone29-92341","DOIUrl":"https://doi.org/10.1115/icone29-92341","url":null,"abstract":"\u0000 10B has a high neutron absorption cross section, so boric acid is generally used as a neutron absorber for reactor control. During an emergency of a nuclear power system, boric acid water is injected into the reactor core by safety injection system to shut down the reactor. Boron concentration has an important impact on the operation safety of the power plant. The mixing process of boric acid and water is influenced by diffusion, convection and turbulence. A new laser-induced fluorescence method is proposed, which can accurately and quickly measure the concentration field of boric acid. This paper introduces the measurement method and the experimental study of boric acid diffusion in T-tube, and compares it with the measurement of wire mesh sensor. In addition, the CFD method is used to calculate the experimental conditions. The results obtained by the three methods were compared with each other, and the diffusion law and research of boric acid are obtained. This study can provide a certain basis for the follow-up study of boron diffusion.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"3 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":"131586419","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}
By studying relevant regulations, standards and good practices, this paper discusses the typical types of common cause failures (CCFs) and general countermeasures in nuclear power plants, and summarizes the methodology, process and defenses for CCFs analysis of safety ventilation and air-conditioning (VAC) system in nuclear power plants. The typical CCFs are divided into four categories, namely functional dependency, spatial dependencies, inherent dependencies and human-related dependencies, and the corresponding defences are functional isolation, physical isolation, and diversity respectively. Comprehensive CCFs risk analysis of VAC system involves plant fault study, equipment FMECA analysis, PSA analysis, hazards analysis and other fields. For the identified CCF points, The optimal solution can be determined by combining factors which include technology maturity, feasibility, engineering cost, etc., and methods such as equipment diversity, physical isolation and functional isolation can be used to reduce or eliminate CCF points, to achieve as low as reasonably practicable (ALARP) risk of CCFs. The CCFs analysis methodology, process and defences summarized in this paper can provide reference for the CCFs of other nuclear safety supporting and auxiliary systems in nuclear power plants.
本文通过对相关法规、标准和良好实践的研究，探讨了核电厂共因故障的典型类型和一般对策，总结了核电厂安全通风空调系统共因故障分析的方法、过程和防御措施。典型的CCFs可分为功能依赖、空间依赖、内在依赖和人相关依赖四类，对应的防御措施分别是功能隔离、物理隔离和多样性隔离。全面的空调系统CCFs风险分析涉及工厂故障研究、设备FMECA分析、PSA分析、危害分析等多个领域。对于已确定的CCF点，可综合技术成熟度、可行性、工程成本等因素确定最优方案，并可采用设备多样性、物理隔离、功能隔离等方法减少或消除CCF点，实现CCF的ALARP (low as reasonable切实可行)风险。本文总结的ccf分析方法、过程和防范措施，可为核电厂其他核安全支撑辅助系统的ccf分析提供参考。
{"title":"The Research on Common Cause Failure Analysis and Countermeasures for Nuclear Safety Ventilation and Air Conditioning System","authors":"Yongsen Peng, Zhengjie Liu","doi":"10.1115/icone29-93557","DOIUrl":"https://doi.org/10.1115/icone29-93557","url":null,"abstract":"\u0000 By studying relevant regulations, standards and good practices, this paper discusses the typical types of common cause failures (CCFs) and general countermeasures in nuclear power plants, and summarizes the methodology, process and defenses for CCFs analysis of safety ventilation and air-conditioning (VAC) system in nuclear power plants. The typical CCFs are divided into four categories, namely functional dependency, spatial dependencies, inherent dependencies and human-related dependencies, and the corresponding defences are functional isolation, physical isolation, and diversity respectively. Comprehensive CCFs risk analysis of VAC system involves plant fault study, equipment FMECA analysis, PSA analysis, hazards analysis and other fields. For the identified CCF points, The optimal solution can be determined by combining factors which include technology maturity, feasibility, engineering cost, etc., and methods such as equipment diversity, physical isolation and functional isolation can be used to reduce or eliminate CCF points, to achieve as low as reasonably practicable (ALARP) risk of CCFs. The CCFs analysis methodology, process and defences summarized in this paper can provide reference for the CCFs of other nuclear safety supporting and auxiliary systems in nuclear power plants.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"69 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":"131694762","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}
Cyber security risk analysis can identify and assess factors that may damage to the system such as digital instrumentation and control system of nuclear power plants. Performing cyber security risk analysis is important for instrumentation and control system of nuclear power plants because it could assess overall impacts of risks and help to identify vulnerabilities to determine next steps to address security risks. With the integration of information system and physical system, cyber security of information system and functional safety of physical system interact with each other, resulting in a type of new comprehensive security problem and introducing serious security risks. Most of the existing cyber security risk analysis methods pay more attention to cyberattacks like attack tree analysis method, Petri net method, and Bayesian network method. STPA-SafeSec is a top-down security risk analysis method focusing on the system itself based on system theory, which starts from unacceptable losses of the system and pays attention to the causal factors that produce unsafe control. In this paper, STPA-SafeSec is applied to the primary circuit pressure control system of high temperature gas-cold reactors in order to perform the hazard analysis of integrated risk assessment for both functional safety and cyber security. The application details are given and a part of the hazardous scenarios tree is obtained for the formulation of mitigation strategies.
{"title":"Integrated Risk Analysis of Function Safety and Cyber Security on I&C System of HTP-PM With STPA-SafeSec","authors":"Yukun Tian, Jianghai Li, Xiaojin Huang","doi":"10.1115/icone29-93395","DOIUrl":"https://doi.org/10.1115/icone29-93395","url":null,"abstract":"\u0000 Cyber security risk analysis can identify and assess factors that may damage to the system such as digital instrumentation and control system of nuclear power plants. Performing cyber security risk analysis is important for instrumentation and control system of nuclear power plants because it could assess overall impacts of risks and help to identify vulnerabilities to determine next steps to address security risks. With the integration of information system and physical system, cyber security of information system and functional safety of physical system interact with each other, resulting in a type of new comprehensive security problem and introducing serious security risks. Most of the existing cyber security risk analysis methods pay more attention to cyberattacks like attack tree analysis method, Petri net method, and Bayesian network method. STPA-SafeSec is a top-down security risk analysis method focusing on the system itself based on system theory, which starts from unacceptable losses of the system and pays attention to the causal factors that produce unsafe control. In this paper, STPA-SafeSec is applied to the primary circuit pressure control system of high temperature gas-cold reactors in order to perform the hazard analysis of integrated risk assessment for both functional safety and cyber security. The application details are given and a part of the hazardous scenarios tree is obtained for the formulation of mitigation strategies.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"96 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":"116105339","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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