Different calculation methods of flow-induced vibration of the reactor vessel internals (RVI) are studied in this paper, that is frequency domain analysis and time domain analysis. Firstly, considering the fluid-structure interaction between the core barrel (CB) and the reactor pressure vessel (RPV), the CB and the heavy reflector, the finite element model of the CB in water is established and the results of its vibration characteristics are obtained, which is consistent with the test results. Based on the flow-induced vibration test of the RVI scale model, the fluctuating pressure data of the CB is obtained. The parameters such as power spectral density (PSD), correlation length and coherence function are obtained by processing the test data. Thus, the cross-power spectral density between any two points is got through these parameters. The root mean square (RMS) response of the CB is obtained by random vibration analysis (frequency domain). Then, the CFD model is established and large eddy simulation (LES) is used to obtain the time history of the fluctuating pressure of the CB. The response of the CB is obtained by time integration method (time domain). The calculation results by using these two methods are in good agreement with the experimental results.
{"title":"Study on Frequency and Time Domain Analysis of Flow-Induced Vibration Response of Core Barrel","authors":"Guitao Cao, He Zhu, Guangdong Liu","doi":"10.1115/icone29-91777","DOIUrl":"https://doi.org/10.1115/icone29-91777","url":null,"abstract":"\u0000 Different calculation methods of flow-induced vibration of the reactor vessel internals (RVI) are studied in this paper, that is frequency domain analysis and time domain analysis. Firstly, considering the fluid-structure interaction between the core barrel (CB) and the reactor pressure vessel (RPV), the CB and the heavy reflector, the finite element model of the CB in water is established and the results of its vibration characteristics are obtained, which is consistent with the test results. Based on the flow-induced vibration test of the RVI scale model, the fluctuating pressure data of the CB is obtained. The parameters such as power spectral density (PSD), correlation length and coherence function are obtained by processing the test data. Thus, the cross-power spectral density between any two points is got through these parameters. The root mean square (RMS) response of the CB is obtained by random vibration analysis (frequency domain). Then, the CFD model is established and large eddy simulation (LES) is used to obtain the time history of the fluctuating pressure of the CB. The response of the CB is obtained by time integration method (time domain). The calculation results by using these two methods are in good agreement with the experimental results.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"134 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":"128796904","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 power plant experience feedback management includes event reporting, screening, analysis, corrective action management and assessment. In the early stage of nuclear power development in China, experience feedback was conducted based on single nuclear power plants. After years of development, the number of nuclear power plants increases year by year, institutes serving multiple nuclear power plants (multi-plants) develop multi-plants experience feedback management system, to improve event management quality, reduce events recurring probability, promote good practice, carry out experience feedback work effectively, reduce the impact of repeated events, continuously improve nuclear safety quality management level, take the implementation of experience feedback and the assessment of its effectiveness as a priority of nuclear safety management. According to the experience feedback assessment of nuclear safety field - operations personnel management, the vulnerabilities of in nuclear safety management have been identified, including insufficient knowledge and substandard operation by operations personnel, and inadequate monitoring of nuclear safety related equipment by operators. The following measures are recommended: improve the on-the-job training outline of operations personnel, and regularly carry out professional knowledge training of machinery, electrical, instrumentation, etc; formulate a special inspection plan for work standardization, and incorporate the special inspection into the task supervision system for tracking management; establish a tracking process for key equipment defects in nuclear safety issues to ensure the timeliness and effectiveness of tracking nuclear safety issues. This article mainly focuses on multi-plants experience feedback organization, information system, program management and nuclear safety indicators, in order to optimize nuclear safety management and prevent the degradation of nuclear safety level.
{"title":"Study on Nuclear Safety Management Based on Multiple Nuclear Power Plants Experience Feedback Management","authors":"Shuang Zhang, Sun Xiaoyan","doi":"10.1115/icone29-93104","DOIUrl":"https://doi.org/10.1115/icone29-93104","url":null,"abstract":"Nuclear power plant experience feedback management includes event reporting, screening, analysis, corrective action management and assessment. In the early stage of nuclear power development in China, experience feedback was conducted based on single nuclear power plants. After years of development, the number of nuclear power plants increases year by year, institutes serving multiple nuclear power plants (multi-plants) develop multi-plants experience feedback management system, to improve event management quality, reduce events recurring probability, promote good practice, carry out experience feedback work effectively, reduce the impact of repeated events, continuously improve nuclear safety quality management level, take the implementation of experience feedback and the assessment of its effectiveness as a priority of nuclear safety management. According to the experience feedback assessment of nuclear safety field - operations personnel management, the vulnerabilities of in nuclear safety management have been identified, including insufficient knowledge and substandard operation by operations personnel, and inadequate monitoring of nuclear safety related equipment by operators. The following measures are recommended: improve the on-the-job training outline of operations personnel, and regularly carry out professional knowledge training of machinery, electrical, instrumentation, etc; formulate a special inspection plan for work standardization, and incorporate the special inspection into the task supervision system for tracking management; establish a tracking process for key equipment defects in nuclear safety issues to ensure the timeliness and effectiveness of tracking nuclear safety issues. This article mainly focuses on multi-plants experience feedback organization, information system, program management and nuclear safety indicators, in order to optimize nuclear safety management and prevent the degradation of nuclear safety level.","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":"134240997","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}
Xingbing Lv, Shuangyang Liu, Hongyang Zhang, Run Lin
Since the external flooding caused by the earthquake and tsunami at Fukushima Daiichi nuclear power plant in Japan, the waterproofing ability of the nuclear power plants, in operation or under construction, has attracted widespread attention from nuclear safety authorities and the public all over the world. While less attention was paid to the conventional island area. The failure of systems and components arranged in the conventional island may lead to the turbine out of the normal operation and impact the operation reliability of the unit, which would influence the generating capacity and availability of the power plant. In this paper, it is selected the combination of “design basis flood (DBF) with the rainfall of a frequency of 10-3/yr” as the analyzed scenario. According to the site characteristics, the variation of external flooding depth on the site over time was calculated under this scenario, the path of external flooding was determined, and the amount of the flooding water, the conventional island ultimately invaded, was calculated. For the availability of the power plant, the corresponding improvement measures were put forward to improve the waterproofing capacity of the plant, which can be a significant reference to the nuclear power plants in our country.
{"title":"Margin Analysis of Conventional Island Under Beyond Design Basis External Flooding Scenario for NPP","authors":"Xingbing Lv, Shuangyang Liu, Hongyang Zhang, Run Lin","doi":"10.1115/icone29-92151","DOIUrl":"https://doi.org/10.1115/icone29-92151","url":null,"abstract":"\u0000 Since the external flooding caused by the earthquake and tsunami at Fukushima Daiichi nuclear power plant in Japan, the waterproofing ability of the nuclear power plants, in operation or under construction, has attracted widespread attention from nuclear safety authorities and the public all over the world. While less attention was paid to the conventional island area. The failure of systems and components arranged in the conventional island may lead to the turbine out of the normal operation and impact the operation reliability of the unit, which would influence the generating capacity and availability of the power plant. In this paper, it is selected the combination of “design basis flood (DBF) with the rainfall of a frequency of 10-3/yr” as the analyzed scenario. According to the site characteristics, the variation of external flooding depth on the site over time was calculated under this scenario, the path of external flooding was determined, and the amount of the flooding water, the conventional island ultimately invaded, was calculated. For the availability of the power plant, the corresponding improvement measures were put forward to improve the waterproofing capacity of the plant, which can be a significant reference to the nuclear power plants in our country.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"53 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":"114269419","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 detection efficiency of soil samples in a cylindrical measuring geometry was calculated using the Monte Carlo method, evaluating the self-absorption corrections in the energy range of 46-2615 keV. By controlling variables, the effects of parameters such as sample density, height and humidity on the self-absorption factor have been studied, and the corresponding correction functions have been obtained. The research results show that: for γ photons of a specific energy, the change of sample density has the greatest impact on self-absorption. For samples whose density is not much different from that of the standard sample, the impact of changes in height and humidity on self-absorption should be considered. In the high-precision measurement of samples containing low-energy γ-photon radionuclides, the errors caused by density, altitude and humidity should be comprehensively considered.
{"title":"Study on Influencing Factors of Soil Sample Self-Absorption With Monte Carlo Method","authors":"Erlei Ye, Chunxia Shen, Hongjie Nan","doi":"10.1115/icone29-92792","DOIUrl":"https://doi.org/10.1115/icone29-92792","url":null,"abstract":"\u0000 The detection efficiency of soil samples in a cylindrical measuring geometry was calculated using the Monte Carlo method, evaluating the self-absorption corrections in the energy range of 46-2615 keV. By controlling variables, the effects of parameters such as sample density, height and humidity on the self-absorption factor have been studied, and the corresponding correction functions have been obtained. The research results show that: for γ photons of a specific energy, the change of sample density has the greatest impact on self-absorption. For samples whose density is not much different from that of the standard sample, the impact of changes in height and humidity on self-absorption should be considered. In the high-precision measurement of samples containing low-energy γ-photon radionuclides, the errors caused by density, altitude and humidity should be comprehensively considered.","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":"116206798","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}
For a more realistic estimation of safety margins, the conservative approach is replaced by integrating the best estimate approach (BE) with uncertainty quantification, the integration which knows as best estimate plus uncertainty (BEPU), which can predict the key safety parameters such as peak cladding temperature (PCT) and departure from nucleate boiling ratio (DNBR), etc. In this sense, a fast and cost-effective tool for uncertainty quantification is developed through a data-driven approach to predict PCT under loss of feedwater accident (LOFW) in AP1000 reactor. This paper includes performing a comparative study between different regression ML algorithms to find the best algorithm which can predict the PCT with higher accuracy. Intent to generate the required data for training and testing the ML algorithm, an uncertainty quantification framework is developed by coupling a best estimate code (RELAP5) with a statistical tool (RAVEN). RELAP5 is used to simulate the thermal-hydraulic response under LOFW accident while a set of uncertainty parameters are propagated through the RELAP5 model using RAVEN. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the RELAP5 code. 5,000 runs were generated in order to acquire a large database for training purposes. The examined algorithms are linear regression, supported vector machine, k-nearest neighbors (KNN), and random forest. The evaluation of algorithms depends mainly on mean absolute error (MAE) and determination coefficient R2. The result shows that the random forest provides high accuracy in predicting PCT within four algorithms, which reaches 98.96%.
{"title":"Comparative Machine Learning Study for Estimating Peak Cladding Temperature in AP1000 Under LOFW","authors":"Merouane Najar, He Wang","doi":"10.1115/icone29-91516","DOIUrl":"https://doi.org/10.1115/icone29-91516","url":null,"abstract":"\u0000 For a more realistic estimation of safety margins, the conservative approach is replaced by integrating the best estimate approach (BE) with uncertainty quantification, the integration which knows as best estimate plus uncertainty (BEPU), which can predict the key safety parameters such as peak cladding temperature (PCT) and departure from nucleate boiling ratio (DNBR), etc. In this sense, a fast and cost-effective tool for uncertainty quantification is developed through a data-driven approach to predict PCT under loss of feedwater accident (LOFW) in AP1000 reactor.\u0000 This paper includes performing a comparative study between different regression ML algorithms to find the best algorithm which can predict the PCT with higher accuracy.\u0000 Intent to generate the required data for training and testing the ML algorithm, an uncertainty quantification framework is developed by coupling a best estimate code (RELAP5) with a statistical tool (RAVEN). RELAP5 is used to simulate the thermal-hydraulic response under LOFW accident while a set of uncertainty parameters are propagated through the RELAP5 model using RAVEN. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the RELAP5 code. 5,000 runs were generated in order to acquire a large database for training purposes. The examined algorithms are linear regression, supported vector machine, k-nearest neighbors (KNN), and random forest. The evaluation of algorithms depends mainly on mean absolute error (MAE) and determination coefficient R2.\u0000 The result shows that the random forest provides high accuracy in predicting PCT within four algorithms, which reaches 98.96%.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"2002 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":"124850730","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-xuan Du, Satya V. Ravikumar Bandaru, W. Villanueva
For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2 × 3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample’s dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25MW/m2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material.
{"title":"Complementary Simulations to Determine Heat Transfer Coefficients and the Maximum Heat Flux in Multi-Nozzle Spray Cooling Experiments","authors":"Yu-xuan Du, Satya V. Ravikumar Bandaru, W. Villanueva","doi":"10.1115/icone29-89086","DOIUrl":"https://doi.org/10.1115/icone29-89086","url":null,"abstract":"\u0000 For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2 × 3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample’s dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25MW/m2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"55 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":"126772927","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}
Under current circumstances, the daily training of the security team in nuclear power plant is carried out by the security company providing security services according to the requirements of the power plant security department, followed by manual assessment and authorization among the guards. There are several problems in this process that can be improved: evaluation is highly subjective, difficult to quantify, and difficult to standardize. Based on the principles of objectivity, unity, comprehensiveness and extendibility, this paper proposes an intelligent assessment method for nuclear power plant security guard capability. Firstly, it decomposes single task into core abilities which are aggregated to an ability library based on their corresponding tasks. Secondly, all task scenarios are made into VR test scenarios with core abilities as the test points, and aggregated into test banks. Then use the VR assessment system for automatically issuing, collecting and scoring test papers. Finally, a comprehensive assessment of the tested guard will be released with the combination of the fuzzy comprehensive assessment method. This method not only can increase objectivity and scientificity of assessment but also has extendibility to new job requirements to meet dynamic changes.
{"title":"Research on Quantitative Assessment Method of Security Guard Capacity in Nuclear Power Plants","authors":"Xiaolin Liu, W. Meng, Jianwen Li, Jinfei Zhang","doi":"10.1115/icone29-90744","DOIUrl":"https://doi.org/10.1115/icone29-90744","url":null,"abstract":"\u0000 Under current circumstances, the daily training of the security team in nuclear power plant is carried out by the security company providing security services according to the requirements of the power plant security department, followed by manual assessment and authorization among the guards. There are several problems in this process that can be improved: evaluation is highly subjective, difficult to quantify, and difficult to standardize. Based on the principles of objectivity, unity, comprehensiveness and extendibility, this paper proposes an intelligent assessment method for nuclear power plant security guard capability. Firstly, it decomposes single task into core abilities which are aggregated to an ability library based on their corresponding tasks. Secondly, all task scenarios are made into VR test scenarios with core abilities as the test points, and aggregated into test banks. Then use the VR assessment system for automatically issuing, collecting and scoring test papers. Finally, a comprehensive assessment of the tested guard will be released with the combination of the fuzzy comprehensive assessment method. This method not only can increase objectivity and scientificity of assessment but also has extendibility to new job requirements to meet dynamic changes.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"29 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":"121212674","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 prestressed containment structure is the last physical barrier against the spread of radioactive material. In addition to being able to resist the pressure caused by internal accidents, it should also have a certain ability to resist external projectiles or explosive shocks. Under the background of frequent terrorist incidents and intensified local conflicts, evaluating the performance of the containment structure under the close-in explosion of explosives carried by drones or light weapons is of great significance for maintaining the safety of nuclear power plants. In this paper, the CEL (Coupled Eulerian and Lagrangian) method is used to calculate the overpressure of the air blast, which is compared with the empirical formula. The test of the damage to the concrete slab under TNT near-in explosion reported in the literature is simulated, and the test results are compared. A numerical model of containment was established and calculated to predict damage behavior. The results show that the air overpressure simulated by the CEL method is in good agreement with the empirical formula, and the simulated results are consistent with the experimental results reported in the literature. Finally, predictions are made for the performance of the containment model under the close-in explosion. This method can be used to design or evaluate the performance of containment against close-in explosions.
{"title":"Analysis of Behavior of Prestressed Containment Under Close-in Explosion Based on CEL Method","authors":"Rongpeng Li, Qinqin Yao, Mengsha Liu, Di Jiang","doi":"10.1115/icone29-92477","DOIUrl":"https://doi.org/10.1115/icone29-92477","url":null,"abstract":"\u0000 The prestressed containment structure is the last physical barrier against the spread of radioactive material. In addition to being able to resist the pressure caused by internal accidents, it should also have a certain ability to resist external projectiles or explosive shocks. Under the background of frequent terrorist incidents and intensified local conflicts, evaluating the performance of the containment structure under the close-in explosion of explosives carried by drones or light weapons is of great significance for maintaining the safety of nuclear power plants. In this paper, the CEL (Coupled Eulerian and Lagrangian) method is used to calculate the overpressure of the air blast, which is compared with the empirical formula. The test of the damage to the concrete slab under TNT near-in explosion reported in the literature is simulated, and the test results are compared. A numerical model of containment was established and calculated to predict damage behavior. The results show that the air overpressure simulated by the CEL method is in good agreement with the empirical formula, and the simulated results are consistent with the experimental results reported in the literature. Finally, predictions are made for the performance of the containment model under the close-in explosion. This method can be used to design or evaluate the performance of containment against close-in explosions.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"18 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":"128408026","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 power plants are important industrial production facilities whose safety issues can have disastrous consequences for the social environment. With the deep fusion of industrialization and information, the nuclear power control system is gradually transformed into automatic digital systems. Such development improves the production efficiency and economic benefits, but introduce the security problem from traditional IT systems into industrial control systems. However, the existing power industrial control system of nuclear power plant mainly adopts firewall, anti-virus software, network gate and other passive security strategies, which lack the safety of active. Besides, only part of the data can be backed up locally and the remote backup is completed manually by operators regularly. The security of industrial control systems cannot be guaranteed and real-time remote backup cannot be realized. In this paper, we analyze and summarize the characteristics of industrial control systems used in nuclear power plants. Then we propose a Truly Private Network (TPN) for nuclear power plants based on the Multi-identifier network (MIN), which integrates blockchain, cryptography, and other security mechanisms to provide a trusted environment. Lastly, we build a network remote real-time backup scheme based on Mimic distributed storage technology providing data tamper-resistance and traceability for nuclear power plants.
{"title":"A MIN-TPN-Based Remote Disaster Recovery and Backup Scheme of Power Industrial Control System in Nuclear Power Plants","authors":"Cui Gang, Xin Yang, Hui Li, Han Wang","doi":"10.1115/icone29-91998","DOIUrl":"https://doi.org/10.1115/icone29-91998","url":null,"abstract":"\u0000 Nuclear power plants are important industrial production facilities whose safety issues can have disastrous consequences for the social environment. With the deep fusion of industrialization and information, the nuclear power control system is gradually transformed into automatic digital systems. Such development improves the production efficiency and economic benefits, but introduce the security problem from traditional IT systems into industrial control systems. However, the existing power industrial control system of nuclear power plant mainly adopts firewall, anti-virus software, network gate and other passive security strategies, which lack the safety of active. Besides, only part of the data can be backed up locally and the remote backup is completed manually by operators regularly. The security of industrial control systems cannot be guaranteed and real-time remote backup cannot be realized. In this paper, we analyze and summarize the characteristics of industrial control systems used in nuclear power plants. Then we propose a Truly Private Network (TPN) for nuclear power plants based on the Multi-identifier network (MIN), which integrates blockchain, cryptography, and other security mechanisms to provide a trusted environment. Lastly, we build a network remote real-time backup scheme based on Mimic distributed storage technology providing data tamper-resistance and traceability for nuclear power plants.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"739 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":"129113868","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}
Li Yang, Xinpeng Li, Yixue Chen, Boxin Wang, Sheng Fang
The Risø Mesoscale PUFF model (RIMPUFF) is a Lagrangian atmospheric dispersion model that uses consecutive Gaussian puffs to simulate accidental release which is widely used in nuclear emergencies. Based on the real-time puff size and the distance between the puff and the monitoring station, RIMPUFF can calculate the dose results quickly by using the dose calculation interpolation table within its dose calculation module but the results are biased. To improve the accuracy of dose results under the requirement of rapid dose calculation, in this paper, a new interpolation table for dose calculation is proposed utilizing a 3D integral dose calculation method. And the proposed interpolation table was evaluated against the Belgian field experiment. In order to assess the quality of the optimization, the dose results of the new interpolation were compared with that of RIMPUFF, the 3D convolution method and observations from the field experiment. The comparisons indicate that the dose results of the new dose interpolation table are closer to the observed values and have better statistical metrics (0.39 for FB, 0.56 for NMSE, 0.72 for FAC2) than that of RIMPUFF (0.53 for FB, 0.76 for NMSE, 0.65 for FAC2).
{"title":"Optimization for Rapid Dose Calculation of the RIMPUFF Model and Its Evaluation Against Belgian Field Experiment","authors":"Li Yang, Xinpeng Li, Yixue Chen, Boxin Wang, Sheng Fang","doi":"10.1115/icone29-90806","DOIUrl":"https://doi.org/10.1115/icone29-90806","url":null,"abstract":"\u0000 The Risø Mesoscale PUFF model (RIMPUFF) is a Lagrangian atmospheric dispersion model that uses consecutive Gaussian puffs to simulate accidental release which is widely used in nuclear emergencies. Based on the real-time puff size and the distance between the puff and the monitoring station, RIMPUFF can calculate the dose results quickly by using the dose calculation interpolation table within its dose calculation module but the results are biased. To improve the accuracy of dose results under the requirement of rapid dose calculation, in this paper, a new interpolation table for dose calculation is proposed utilizing a 3D integral dose calculation method. And the proposed interpolation table was evaluated against the Belgian field experiment. In order to assess the quality of the optimization, the dose results of the new interpolation were compared with that of RIMPUFF, the 3D convolution method and observations from the field experiment. The comparisons indicate that the dose results of the new dose interpolation table are closer to the observed values and have better statistical metrics (0.39 for FB, 0.56 for NMSE, 0.72 for FAC2) than that of RIMPUFF (0.53 for FB, 0.76 for NMSE, 0.65 for FAC2).","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"43 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":"115857049","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：