Huian Zhan, Shaokun Jiang, Zhiyong Tao, Ning Zhao, Shengnan Yuan
The safety control of hydrogen is a key safety issue that must be considered in the design of nuclear power plant. In order to better study the impact of hydrogen combustion and even deflagration on the nuclear power plant, the Handan Purification Equipment Research Institute has designed and built a hydrogen deflagration environment simulation test platform, which can simulate the deflagration process of mixed gas with high hydrogen concentration range. The test results show that there are some differences in the hydrogen deflagration process of test container with different volumes. The large volume test container is more in line with the change process of hydrogen deflagration environment in serious accidents of nuclear power plant. The large volume test container has problems such as large concentration gradient of hydrogen and temperature stratification. The closer to the top of the test container, the higher the temperature. It is possible to select a suitable location to install the test equipment according to the needs. With the same gas composition conditions, increasing the initial mixture temperature can effectively improve the half peak width of the hydrogen deflagration ambient temperature curve and slow down the temperature decline rate in the simulated environment. The above experimental research lays a foundation for the equipment availability verification test in the hydrogen deflagration environment of nuclear power plant.
{"title":"Study on Simulation Test of Hydrogen Deflagration Environment In Nuclear Power Plants","authors":"Huian Zhan, Shaokun Jiang, Zhiyong Tao, Ning Zhao, Shengnan Yuan","doi":"10.1115/icone29-92375","DOIUrl":"https://doi.org/10.1115/icone29-92375","url":null,"abstract":"\u0000 The safety control of hydrogen is a key safety issue that must be considered in the design of nuclear power plant. In order to better study the impact of hydrogen combustion and even deflagration on the nuclear power plant, the Handan Purification Equipment Research Institute has designed and built a hydrogen deflagration environment simulation test platform, which can simulate the deflagration process of mixed gas with high hydrogen concentration range. The test results show that there are some differences in the hydrogen deflagration process of test container with different volumes. The large volume test container is more in line with the change process of hydrogen deflagration environment in serious accidents of nuclear power plant. The large volume test container has problems such as large concentration gradient of hydrogen and temperature stratification. The closer to the top of the test container, the higher the temperature. It is possible to select a suitable location to install the test equipment according to the needs. With the same gas composition conditions, increasing the initial mixture temperature can effectively improve the half peak width of the hydrogen deflagration ambient temperature curve and slow down the temperature decline rate in the simulated environment. The above experimental research lays a foundation for the equipment availability verification test in the hydrogen deflagration environment of nuclear power plant.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"46 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":"125340060","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}
Due to the perspective ability of the internal structure of objects, the X-ray security screening system has been used to ensure nuclear security in nuclear facility entrancement. To improve detection efficiency and accuracy, recent work posed a particular interest in automated threat detection algorithms based on deep learning. A significant obstacle to developing high performance of such algorithms is the difficulty of obtaining large labeled training datasets. X-ray image data augmentation strategies based on experimental images have been proposed by other investigators. This paper proposed a physics-driven X-ray image data augmentation method for automated threat detection. Using the 3D modeling software, we can construct threat models and project them into threat images without experiment. According to Lambert-Beer law, the simulation image datasets can be made by exponentially overlaying the threat images on the background images. Considering the energy spectrum and scattering effect, we further process the phantoms and projection images accordingly to be more authentic. We train YOLOv5 architecture on the simulation dataset and test the algorithm on the experimental images. The results show that our approach achieves good performances in automated threat detection with an average recognition accuracy of over 90% and a mAP@0.5 of 82.9%, which is an effective method to increase the X-ray image dataset in both quantity and diversity.
{"title":"A Physics-Driven X-Ray Image Data Augmentation Method for Automated Threat Detection in Nuclear Facility Entrancement","authors":"Shuo Xu, Gang Chen, Weiwei Li, Xincheng Xiang","doi":"10.1115/icone29-92402","DOIUrl":"https://doi.org/10.1115/icone29-92402","url":null,"abstract":"\u0000 Due to the perspective ability of the internal structure of objects, the X-ray security screening system has been used to ensure nuclear security in nuclear facility entrancement. To improve detection efficiency and accuracy, recent work posed a particular interest in automated threat detection algorithms based on deep learning. A significant obstacle to developing high performance of such algorithms is the difficulty of obtaining large labeled training datasets. X-ray image data augmentation strategies based on experimental images have been proposed by other investigators. This paper proposed a physics-driven X-ray image data augmentation method for automated threat detection. Using the 3D modeling software, we can construct threat models and project them into threat images without experiment. According to Lambert-Beer law, the simulation image datasets can be made by exponentially overlaying the threat images on the background images. Considering the energy spectrum and scattering effect, we further process the phantoms and projection images accordingly to be more authentic. We train YOLOv5 architecture on the simulation dataset and test the algorithm on the experimental images. The results show that our approach achieves good performances in automated threat detection with an average recognition accuracy of over 90% and a mAP@0.5 of 82.9%, which is an effective method to increase the X-ray image dataset in both quantity and diversity.","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":"125564770","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}
Wanxin Feng, Runcai Zhao, Shuxian Ding, Yu Yu, F. Niu
At present, the probabilistic safety assessment (PSA) method is mainly applied to a single unit, while actual nuclear power plants are usually multi-reactor sites. In the multi-unit level 1 probabilistic safety assessment study, the modeling workload is greatly increased. Also, it is necessary to consider the correlation between units, the probabilistic safety evaluation of multiple reactor sites is confronted with great challenges. Safety assessment of multi-unit sites should begin with a screening of the initial event, and not all events need to be analyzed by building an event tree. This paper described the classification status of initial events at multi-unit and calculated these events’ core damage frequency (CDF) upper and lower bounds. The upper limit of CDF is based on the assumption of complete dependence between units, and the lower limit is based on the assumption of complete independence between units. Taking a two-unit as an example, it screened the initial events and sorted out the characteristics of accidents. We assume that the time limit for simultaneous accidents of two units is 24 hours. The results show that the loss-of-offsite-power (LOOP), loss of heat trap and other events are suitable for modeling and analysis. In addition, given the analysis suggestions for other accidents combined with their characteristics. The analytical methods and results provide the basis for the multi-unit Probabilistic Safety Assessment modeling.
{"title":"Multi-Units Nuclear Power Plant Site Initial Event Filtration","authors":"Wanxin Feng, Runcai Zhao, Shuxian Ding, Yu Yu, F. Niu","doi":"10.1115/icone29-94324","DOIUrl":"https://doi.org/10.1115/icone29-94324","url":null,"abstract":"\u0000 At present, the probabilistic safety assessment (PSA) method is mainly applied to a single unit, while actual nuclear power plants are usually multi-reactor sites. In the multi-unit level 1 probabilistic safety assessment study, the modeling workload is greatly increased. Also, it is necessary to consider the correlation between units, the probabilistic safety evaluation of multiple reactor sites is confronted with great challenges. Safety assessment of multi-unit sites should begin with a screening of the initial event, and not all events need to be analyzed by building an event tree. This paper described the classification status of initial events at multi-unit and calculated these events’ core damage frequency (CDF) upper and lower bounds. The upper limit of CDF is based on the assumption of complete dependence between units, and the lower limit is based on the assumption of complete independence between units. Taking a two-unit as an example, it screened the initial events and sorted out the characteristics of accidents. We assume that the time limit for simultaneous accidents of two units is 24 hours. The results show that the loss-of-offsite-power (LOOP), loss of heat trap and other events are suitable for modeling and analysis. In addition, given the analysis suggestions for other accidents combined with their characteristics. The analytical methods and results provide the basis for the multi-unit Probabilistic Safety Assessment modeling.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"30 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":"114271164","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}
Qian Sun, Zhipeng Wang, Changwu Wang, Yuhang Zhang, Lei Chen, Xuexin Wang, Guoqiang Li
For radioactive material packages railway transportation, a reliable fastening system of the packages in the wagon is the key to realize safe and efficient transportation. The following types of fastening systems are mainly used in the railway transportation for small radioactive material packages, such as stopper reinforcement, rope reinforcement, bolt reinforcement, freight container loading and reinforcement, etc. The fastening system should be determined according to the size, shape, weight and other parameters of different packages. In that case, the requirements of convenient operation could be considered while ensuring the reinforcement strength. Combined with the examples of different types of radioactive material packages, some examples of reasonable fastening systems for railway transportation were introduced. In order to ensure the safety, the strength of reinforcement materials should be calculated and verified, and the transportation stability of wagon loaded with packages should be analyzed.
{"title":"Analysis on the Fastening System of Radioactive Material Packages In Railway Transportation","authors":"Qian Sun, Zhipeng Wang, Changwu Wang, Yuhang Zhang, Lei Chen, Xuexin Wang, Guoqiang Li","doi":"10.1115/icone29-90719","DOIUrl":"https://doi.org/10.1115/icone29-90719","url":null,"abstract":"\u0000 For radioactive material packages railway transportation, a reliable fastening system of the packages in the wagon is the key to realize safe and efficient transportation. The following types of fastening systems are mainly used in the railway transportation for small radioactive material packages, such as stopper reinforcement, rope reinforcement, bolt reinforcement, freight container loading and reinforcement, etc. The fastening system should be determined according to the size, shape, weight and other parameters of different packages. In that case, the requirements of convenient operation could be considered while ensuring the reinforcement strength. Combined with the examples of different types of radioactive material packages, some examples of reasonable fastening systems for railway transportation were introduced. In order to ensure the safety, the strength of reinforcement materials should be calculated and verified, and the transportation stability of wagon loaded with packages should be analyzed.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"47 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":"124864431","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}
Xiaocong Deng, Qingsong Liu, Jianming Zhou, Chen Xi, Shuo Hou, Qiang Liu, G. Lu, Xudong Wei, Liang Ni
For the large high-temperature structure with multiple points of support, the support keys are not only subjected to the self-weight of the structure, but also to the static and dynamic loads in the assembly environment; on the other hand, the nonuniform temperature distribution forms different degrees of thermal stresses on the structure, while the thermal expansion effect has a significant impact on the assembly state of the support keys and the supports. As an over-constrained system, if the extreme assembly relationship between the bearing key and the support occurs, such as jamming, it will further deteriorate the pre-stress condition of the support system and seriously threaten the safety of the structure. Restricted by the size of the structure and test conditions, the scaled model is considered instead of the full-sized structure for dynamics analysis and experimental study to evaluate the safety and reliability of the support system under thermal environment. In this paper, the force state of the bearing system under operation is analyzed by combining numerical simulation and scaled model tests, the possible jamming risk during the application of the bearing is predicted and the corresponding plan is formulated. The effectiveness of the plan is verified by optimizing the clearance factor of the scaled model, which provides a reference for the safe design of multi-point support structures under thermal environment.
{"title":"Dynamic Analysis and Experimental Study of the Multi-Point Support Structure Based on Similar Theory Under Thermal Environment","authors":"Xiaocong Deng, Qingsong Liu, Jianming Zhou, Chen Xi, Shuo Hou, Qiang Liu, G. Lu, Xudong Wei, Liang Ni","doi":"10.1115/icone29-93497","DOIUrl":"https://doi.org/10.1115/icone29-93497","url":null,"abstract":"\u0000 For the large high-temperature structure with multiple points of support, the support keys are not only subjected to the self-weight of the structure, but also to the static and dynamic loads in the assembly environment; on the other hand, the nonuniform temperature distribution forms different degrees of thermal stresses on the structure, while the thermal expansion effect has a significant impact on the assembly state of the support keys and the supports. As an over-constrained system, if the extreme assembly relationship between the bearing key and the support occurs, such as jamming, it will further deteriorate the pre-stress condition of the support system and seriously threaten the safety of the structure. Restricted by the size of the structure and test conditions, the scaled model is considered instead of the full-sized structure for dynamics analysis and experimental study to evaluate the safety and reliability of the support system under thermal environment. In this paper, the force state of the bearing system under operation is analyzed by combining numerical simulation and scaled model tests, the possible jamming risk during the application of the bearing is predicted and the corresponding plan is formulated. The effectiveness of the plan is verified by optimizing the clearance factor of the scaled model, which provides a reference for the safe design of multi-point support structures under thermal environment.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"27 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":"127834798","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}
Real-time and automatic radioisotope identification using gamma spectrum is an important issue in the field of nuclear safety. It is widely used in vehicle mounted radioisotope monitoring, Marine radioisotope monitoring and nuclear decommissioning verification scenarios. At present, the focus of radionuclide identification is fast and stable recognition under low count conditions. In this paper, a radionuclide recognition method based explainable artificial neural network is proposed, and a synthetic gamma spectrum data set is created. The data set contains gamma-ray spectra of 12 different types of radionuclides, which were obtained by Geant4 Monte Carlo simulation software and gaussian broadening of the detector. Data Augmentation was achieved by simulating gamma spectra at different measuring times, different measuring distances and different ambient temperatures. The training results of the neural network optimized by hyperparameter show that it has a high accuracy on the test set with shorter measurement time, longer measurement distance and larger energy spectrum drift range, which provides a method for rapid identification of nuclides in the case of low count. Using t-SNE dimension reduction technology, the twelve dimensions data output by the neural network is reduced to two dimensions for feature visualization, which vividly explains and verifies the recognition results of neural network.
{"title":"Explainable Neural Network Algorithm for Rapid Radionuclide Identification Under Low Count Gamma-Ray Spectrum Data","authors":"Yu Wang, Sufen Li, Yong-gang Huo, Jianqing Yang, Quan-hu Zhang","doi":"10.1115/icone29-92829","DOIUrl":"https://doi.org/10.1115/icone29-92829","url":null,"abstract":"\u0000 Real-time and automatic radioisotope identification using gamma spectrum is an important issue in the field of nuclear safety. It is widely used in vehicle mounted radioisotope monitoring, Marine radioisotope monitoring and nuclear decommissioning verification scenarios. At present, the focus of radionuclide identification is fast and stable recognition under low count conditions. In this paper, a radionuclide recognition method based explainable artificial neural network is proposed, and a synthetic gamma spectrum data set is created. The data set contains gamma-ray spectra of 12 different types of radionuclides, which were obtained by Geant4 Monte Carlo simulation software and gaussian broadening of the detector. Data Augmentation was achieved by simulating gamma spectra at different measuring times, different measuring distances and different ambient temperatures. The training results of the neural network optimized by hyperparameter show that it has a high accuracy on the test set with shorter measurement time, longer measurement distance and larger energy spectrum drift range, which provides a method for rapid identification of nuclides in the case of low count. Using t-SNE dimension reduction technology, the twelve dimensions data output by the neural network is reduced to two dimensions for feature visualization, which vividly explains and verifies the recognition results of neural network.","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":"126904307","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In the view of the International Atomic Energy Agency (IAEA), the decision-making of safety supervision of civil nuclear facilities needs independence. In order to deeply evaluate and review a series of issues related to the safety of civil nuclear facilities, national regulatory authorities usually need to consult independent institutions or external committees with professional capabilities. Such organizations providing technical services are called Technical Support Organizations (TSO), which can be national laboratories, research institutions or consulting units. In some cases, support from professional organizations in other countries may also be required. Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. All along, China has continuously strengthened the construction of nuclear safety supervision capacity, and the safety supervision capacity has been gradually improved. China has established a nuclear and radiation safety supervision system, including National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional nuclear and radiation safety supervision stations, Nuclear and Radiation Safety Center (NSC) and other units. China has established a nuclear safety expert committee, and mainly relies on the National Nuclear and Radiation Safety Supervision Technology R & D Base (R & D Base) as the central level test, verification and R & D platform to provide consultation and technical support. The R & D base has completed the construction of some test, verification and R & D functions. China has built a series of important laboratories in the R & D base, it still has not formed R & D system. In the future, China need to improve the technical support capacity of nuclear safety supervision. In short, the independence of nuclear safety regulatory decision-making requires to enhance from institutional independence to capability independence. The independence of China’s nuclear safety regulatory authority has been basically established. In the future, it is more important to improve the regulatory capacity. Only by establishing a perfect technology R & D system can nuclear safety decision-making be based on sufficient scientific basis.
{"title":"Analysis on the Nuclear Safety Supervision Mode of the World’s Major Nuclear Power Countries and Its Enlightenment to the Improvement of China’s Nuclear Safety Supervision Technical Support Ability","authors":"Han Wu, Guoxin Yu, K. Teng, Xiangyang Zheng","doi":"10.1115/icone29-91379","DOIUrl":"https://doi.org/10.1115/icone29-91379","url":null,"abstract":"\u0000 In the view of the International Atomic Energy Agency (IAEA), the decision-making of safety supervision of civil nuclear facilities needs independence. In order to deeply evaluate and review a series of issues related to the safety of civil nuclear facilities, national regulatory authorities usually need to consult independent institutions or external committees with professional capabilities. Such organizations providing technical services are called Technical Support Organizations (TSO), which can be national laboratories, research institutions or consulting units. In some cases, support from professional organizations in other countries may also be required.\u0000 Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. All along, China has continuously strengthened the construction of nuclear safety supervision capacity, and the safety supervision capacity has been gradually improved.\u0000 China has established a nuclear and radiation safety supervision system, including National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional nuclear and radiation safety supervision stations, Nuclear and Radiation Safety Center (NSC) and other units. China has established a nuclear safety expert committee, and mainly relies on the National Nuclear and Radiation Safety Supervision Technology R & D Base (R & D Base) as the central level test, verification and R & D platform to provide consultation and technical support.\u0000 The R & D base has completed the construction of some test, verification and R & D functions. China has built a series of important laboratories in the R & D base, it still has not formed R & D system. In the future, China need to improve the technical support capacity of nuclear safety supervision.\u0000 In short, the independence of nuclear safety regulatory decision-making requires to enhance from institutional independence to capability independence. The independence of China’s nuclear safety regulatory authority has been basically established. In the future, it is more important to improve the regulatory capacity. Only by establishing a perfect technology R & D system can nuclear safety decision-making be based on sufficient scientific basis.","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":"126106606","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}
Passive containment air cooling system (PAS) is an important special safety system for the small modular pressurized water reactor ACP-100. The heat is transferred to the environment by the natural draft air cooling under accident conditions. The flow loss of the air flow paths has an important influence on the heat transfer power of PAS. A one-sixth scale model of PAS air flow path was constructed which was a 90° wedge angle sector of the cylindrical symmetric passages. The pressure distributions at different sections of the model were measured, the drag coefficients were analyzed. The experimental results showed that the total drag coefficients of the four different air inlet paths were almost the same, and they showed a decreasing trend with the increase of Reynolds number. The sleeves had the largest drag coefficient, accounting for about 26%, followed by the air inlet, chimney outlet, the turning and the corridor. The other components had little effect on the pressure loss. Afterward, the aerodynamic modification was carried out, the modification for the turning could reduce the total drag coefficient by 4.5 %. The experimental results of this paper provide necessary data support for the ACP 100 engineering design.
{"title":"Experimental Research on Drag Characteristic of Air Flow Path for ACP-100 Passive Containment Air Cooling System","authors":"Mingrui Yu, Hongliang Wang, Zhuo Liu, Yu Feng, Xu Han, Yidan Yuan","doi":"10.1115/icone29-91504","DOIUrl":"https://doi.org/10.1115/icone29-91504","url":null,"abstract":"\u0000 Passive containment air cooling system (PAS) is an important special safety system for the small modular pressurized water reactor ACP-100. The heat is transferred to the environment by the natural draft air cooling under accident conditions. The flow loss of the air flow paths has an important influence on the heat transfer power of PAS. A one-sixth scale model of PAS air flow path was constructed which was a 90° wedge angle sector of the cylindrical symmetric passages. The pressure distributions at different sections of the model were measured, the drag coefficients were analyzed. The experimental results showed that the total drag coefficients of the four different air inlet paths were almost the same, and they showed a decreasing trend with the increase of Reynolds number. The sleeves had the largest drag coefficient, accounting for about 26%, followed by the air inlet, chimney outlet, the turning and the corridor. The other components had little effect on the pressure loss. Afterward, the aerodynamic modification was carried out, the modification for the turning could reduce the total drag coefficient by 4.5 %. The experimental results of this paper provide necessary data support for the ACP 100 engineering design.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"37 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":"125574916","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}
After the high-energy pipeline break accident in the containment, the pipeline insulation and other materials near the break will be damaged after impact, which will produce a large amount of debris. Some debris will be produced, a fraction of this dislodged insulation and other materials will be transported to the containment floor by the steam/water flows from the break and the containment spray. The in-containment refueling water storage tank is widely used in the third-generation nuclear power plant as the water source to support long-term recirculation for the functions of residual heat removal, emergency core cooling, and containment atmosphere cleanup. Typically, a filter is installed around inlet of pump to filter the debris and to minimize the amount of debris entering the downstream lines and the reactor core. The accumulated debris on the screen of filter may increase the differential pressure across the screen and thus decrease the net positive suction head margin available to pumps. The system may even disable all operational capability. In order to evaluating the performance of filters during and after accident, a reasonable conservative debris evaluation is required as design input. In order to accurately evaluate the impact of filter after accident, a reasonable and conservative debris source term evaluation is needed as the design input. The nuclear island design of the third-generation nuclear power first project fully adopts 3D modeling. This paper introduces the improvement of the upstream analysis method of IRWST filter based on three-dimensional design.
{"title":"Study on In-Containment Refueling Water Storage Tank Filter Upstream Effects Analysis Based on Three-Dimensional Design","authors":"Minghua Zhu","doi":"10.1115/icone29-93200","DOIUrl":"https://doi.org/10.1115/icone29-93200","url":null,"abstract":"\u0000 After the high-energy pipeline break accident in the containment, the pipeline insulation and other materials near the break will be damaged after impact, which will produce a large amount of debris. Some debris will be produced, a fraction of this dislodged insulation and other materials will be transported to the containment floor by the steam/water flows from the break and the containment spray. The in-containment refueling water storage tank is widely used in the third-generation nuclear power plant as the water source to support long-term recirculation for the functions of residual heat removal, emergency core cooling, and containment atmosphere cleanup. Typically, a filter is installed around inlet of pump to filter the debris and to minimize the amount of debris entering the downstream lines and the reactor core. The accumulated debris on the screen of filter may increase the differential pressure across the screen and thus decrease the net positive suction head margin available to pumps. The system may even disable all operational capability. In order to evaluating the performance of filters during and after accident, a reasonable conservative debris evaluation is required as design input. In order to accurately evaluate the impact of filter after accident, a reasonable and conservative debris source term evaluation is needed as the design input. The nuclear island design of the third-generation nuclear power first project fully adopts 3D modeling. This paper introduces the improvement of the upstream analysis method of IRWST filter based on three-dimensional design.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"122 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":"128145768","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}
Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen volume fraction can be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. The measurement accuracy of hydrogen volume fraction in containment is important to judge the flammability of hydrogen in the containment during a severe accident. The experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes will be set up in the future, it is necessary to study on the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes. Computational fluid dynamics (CFD) codes can be applied to predict hydrogen distribution in compartments at steady state in the containment during a severe accident. The pressure vessel is used to provide the same mixed gas as in the containment during a severe accident, and the numerical simulation is carried out for different work conditions with hydrogen-nitrogen mixture of 20% hydrogen volume fraction in the pressure vessel, investigating the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes. The numerical simulation results are qualitatively analyzed in the paper, the following conclusions indicate that it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0–100, 0–120, 0–140, 0–160) with different mixed gas temperatures and it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0.1–160, 1–160) with different steam injection rates. Therefore, mixed gas temperature and steam injection rate have almost no effect on hydrogen distribution at steady state in small volumes. The numerical simulation results can provide some data basis for the experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes in the future, and the outcomes of the qualitatively analysis can be used for design and optimization of the experimental device, in order to develop a hydrogen volume fraction measurement system for the containment during a severe accident.
{"title":"Numerical Simulation of the Effect of Mixed Gas Temperature and Steam Injection Rate on Hydrogen Distribution at Steady State in Small Volumes","authors":"Yu Feng, Hongliang Wang, Mingrui Yu, Yidan Yuan","doi":"10.1115/icone29-91506","DOIUrl":"https://doi.org/10.1115/icone29-91506","url":null,"abstract":"\u0000 Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen volume fraction can be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. The measurement accuracy of hydrogen volume fraction in containment is important to judge the flammability of hydrogen in the containment during a severe accident. The experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes will be set up in the future, it is necessary to study on the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes.\u0000 Computational fluid dynamics (CFD) codes can be applied to predict hydrogen distribution in compartments at steady state in the containment during a severe accident. The pressure vessel is used to provide the same mixed gas as in the containment during a severe accident, and the numerical simulation is carried out for different work conditions with hydrogen-nitrogen mixture of 20% hydrogen volume fraction in the pressure vessel, investigating the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes. The numerical simulation results are qualitatively analyzed in the paper, the following conclusions indicate that it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0–100, 0–120, 0–140, 0–160) with different mixed gas temperatures and it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0.1–160, 1–160) with different steam injection rates. Therefore, mixed gas temperature and steam injection rate have almost no effect on hydrogen distribution at steady state in small volumes.\u0000 The numerical simulation results can provide some data basis for the experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes in the future, and the outcomes of the qualitatively analysis can be used for design and optimization of the experimental device, in order to develop a hydrogen volume fraction measurement system for the containment during a severe accident.","PeriodicalId":365848,"journal":{"name":"Volume 5: Nuclear Safety, Security, and Cyber Security","volume":"15 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":"128018930","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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