Xiang Li, Wan Sun, Zhuqing Wen, Luteng Zhang, Zaiyong Ma, Liang-ming Pan
In the event of a Loss of Coolant Accident, a large amount of coolant is released from the reactor, and steam is generated inside the reactor, and then flows down the heat pipe into the steam generator. When a natural circulation is not established in a loop, the condensed liquid in the steam generator is returned to the core by gravity of the wall of the pipe through the hot leg. This countercurrent flow has important influence on reactor safety analysis. At present, many researches have been carried out in this field, but most of them focus on the study of the countercurrent flow characteristics of two-phase flow in vertical or horizontal pipes, there are few researches on the countercurrent flow in a complex. In this paper, the countercurrent flow of two-phase flow in the hot leg under the condition of Loss of Coolant Accident (LOCA) is modeled and calculated. The experimental loop is composed of vertical pipe, inclined pipe and horizontal pipe. It is found that the shape of the hot leg has a significant effect on the injection limit, and the data based on dimensionless square root of velocity is compared with Wallis’ vertical tube results. The effects of liquid viscosity, density of liquid phase and gas, and geometry of pipeline on flooding were also studied.
{"title":"Countercurrent Flow Limitation Calculation of Hot Leg Based On System Program","authors":"Xiang Li, Wan Sun, Zhuqing Wen, Luteng Zhang, Zaiyong Ma, Liang-ming Pan","doi":"10.1115/icone29-90971","DOIUrl":"https://doi.org/10.1115/icone29-90971","url":null,"abstract":"\u0000 In the event of a Loss of Coolant Accident, a large amount of coolant is released from the reactor, and steam is generated inside the reactor, and then flows down the heat pipe into the steam generator. When a natural circulation is not established in a loop, the condensed liquid in the steam generator is returned to the core by gravity of the wall of the pipe through the hot leg. This countercurrent flow has important influence on reactor safety analysis. At present, many researches have been carried out in this field, but most of them focus on the study of the countercurrent flow characteristics of two-phase flow in vertical or horizontal pipes, there are few researches on the countercurrent flow in a complex. In this paper, the countercurrent flow of two-phase flow in the hot leg under the condition of Loss of Coolant Accident (LOCA) is modeled and calculated. The experimental loop is composed of vertical pipe, inclined pipe and horizontal pipe. It is found that the shape of the hot leg has a significant effect on the injection limit, and the data based on dimensionless square root of velocity is compared with Wallis’ vertical tube results. The effects of liquid viscosity, density of liquid phase and gas, and geometry of pipeline on flooding were also studied.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"12 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":"114628787","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}
Concerning the Fukushima nuclear accident, many radionuclides were released into the marine environment, which caused contamination of many parts of the world through ocean circulation. Regarding inland nuclear power plants, freshwater habitats such as reservoirs and rivers could also be polluted by the radioactive effluents. Several models were developed to track radionuclide transport in the water environment. However, the applicability and characteristics of these models were not fully identified and compared, particularly for the different water environments. In this study, the research progress and application examples of radionuclide transport models in rivers and oceans in recent years are systematically summarized, and the methods for simulating radionuclide transport behavior in various water environments are compared. An adequate model is expected to be instrumental in assessing radioactive pollution in various water environments.
{"title":"Investigation of Transport Models of Radionuclides in Various Water Environments","authors":"Zhengzhe Qu, F. Xie, Baojie Nie, Liang Wang","doi":"10.1115/icone29-88889","DOIUrl":"https://doi.org/10.1115/icone29-88889","url":null,"abstract":"\u0000 Concerning the Fukushima nuclear accident, many radionuclides were released into the marine environment, which caused contamination of many parts of the world through ocean circulation. Regarding inland nuclear power plants, freshwater habitats such as reservoirs and rivers could also be polluted by the radioactive effluents. Several models were developed to track radionuclide transport in the water environment. However, the applicability and characteristics of these models were not fully identified and compared, particularly for the different water environments. In this study, the research progress and application examples of radionuclide transport models in rivers and oceans in recent years are systematically summarized, and the methods for simulating radionuclide transport behavior in various water environments are compared. An adequate model is expected to be instrumental in assessing radioactive pollution in various water environments.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"354 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":"123212209","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}
Zilin Zhou, Yu Wang, Jingni Guo, F. Xie, Wenqian Li, Y. Wen, B. Shan
Tritium is an important nuclear fuel in fusion reactors and one of the dominant nuclides in the primary coolant in fission reactors. The storage, feeding, control, monitoring, and transport of tritium are important for practical engineering applications. Because of the high mobility of tritium in both fission and fusion nuclear systems and its effect on human body, tritium has received great attention worldwide. The retention and prevention of tritium permeation from primary loop to secondary loop is a common research interest in various reactors. Previous studies indicated that a surface oxide layer is an efficient method to reduce tritium permeation. In this paper, we summarize the effects of various oxide layers on the permeation of hydrogen isotopes in nuclear reactors. The permeation reduction factor for materials used in fusion reactors ranges from 1000 to 100000. The diffusion behavior of tritium in several materials with and without oxide layer is discussed in detail. The oxide layer is more important than intrinsic permeability to prevent tritium permeation. As tritium is the only radioactive source term in the secondary loop of high temperature gas-cooled reactors, we also reviewed the permeation of hydrogen isotopes in the heat exchanger, which is an important issue in nuclear hydrogen production. The present study provides a comprehensive overview of tritium permeation behavior and is expected to promote the development and design of tritium-permeation-proof materials in the future.
{"title":"Summary of Effects of Oxide Layer on Permeability Coefficient of Tritium","authors":"Zilin Zhou, Yu Wang, Jingni Guo, F. Xie, Wenqian Li, Y. Wen, B. Shan","doi":"10.1115/icone29-91275","DOIUrl":"https://doi.org/10.1115/icone29-91275","url":null,"abstract":"\u0000 Tritium is an important nuclear fuel in fusion reactors and one of the dominant nuclides in the primary coolant in fission reactors. The storage, feeding, control, monitoring, and transport of tritium are important for practical engineering applications. Because of the high mobility of tritium in both fission and fusion nuclear systems and its effect on human body, tritium has received great attention worldwide. The retention and prevention of tritium permeation from primary loop to secondary loop is a common research interest in various reactors. Previous studies indicated that a surface oxide layer is an efficient method to reduce tritium permeation. In this paper, we summarize the effects of various oxide layers on the permeation of hydrogen isotopes in nuclear reactors. The permeation reduction factor for materials used in fusion reactors ranges from 1000 to 100000. The diffusion behavior of tritium in several materials with and without oxide layer is discussed in detail. The oxide layer is more important than intrinsic permeability to prevent tritium permeation. As tritium is the only radioactive source term in the secondary loop of high temperature gas-cooled reactors, we also reviewed the permeation of hydrogen isotopes in the heat exchanger, which is an important issue in nuclear hydrogen production. The present study provides a comprehensive overview of tritium permeation behavior and is expected to promote the development and design of tritium-permeation-proof materials in the future.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","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":"129870071","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}
Matrix graphite (MG) constitutes a major component of the fuel element in high-temperature gas-cooled reactors (HTRs), and its performance directly affects the economy and safety of HTRs. A3-3 MG, developed in Germany in the last century, is still used today without significant changes. However, with the successful commercialization of HTR, there is an urgent need to develop MG that is suitable for the characteristics of each country. In this paper, comparative studies on the powder processing of natural flake graphite were carried out based on the raw materials from Inner Mongolia. First, the influence of the sequence of purification and pulverization on the basic properties of graphite powder was studied. Second, the influence of the degree of spheroidization of the natural flake graphite powder on the properties of the MG was further investigated, and the controlling mechanism of the spheroidization process on the microstructure and physical properties of the MG was revealed. The results of this study provide important insights for optimizing the preparation process of HTR fuel element MG.
{"title":"Pulverization and Spheroidization of Natural Graphite Powder for The Matrix Graphite in High-Temperature Gas-Cooled Reactors","authors":"X. Cao, Zhen Lu, Yuxiao Feng, Juan Bai, Bing Liu, Keya Shen","doi":"10.1115/icone29-92524","DOIUrl":"https://doi.org/10.1115/icone29-92524","url":null,"abstract":"\u0000 Matrix graphite (MG) constitutes a major component of the fuel element in high-temperature gas-cooled reactors (HTRs), and its performance directly affects the economy and safety of HTRs. A3-3 MG, developed in Germany in the last century, is still used today without significant changes. However, with the successful commercialization of HTR, there is an urgent need to develop MG that is suitable for the characteristics of each country. In this paper, comparative studies on the powder processing of natural flake graphite were carried out based on the raw materials from Inner Mongolia. First, the influence of the sequence of purification and pulverization on the basic properties of graphite powder was studied. Second, the influence of the degree of spheroidization of the natural flake graphite powder on the properties of the MG was further investigated, and the controlling mechanism of the spheroidization process on the microstructure and physical properties of the MG was revealed. The results of this study provide important insights for optimizing the preparation process of HTR fuel element MG.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","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":"125410591","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}
Sina Li, Binhuo Yan, Guojun Tang, Donghong Peng, Hongzhi Lu
Traditional heat pipe cooled reactors have the characteristics of high core temperature and high fuel enrichment, which makes its development limited by several unfavorable factors, such as the limitation of high temperature materials and low burnup, etc. In this paper, a new heat pipe cooled reactor with a rated thermal power of 1.5MWt is proposed to address the problems of high operating temperature and high fuel enrichment. Low-enriched uranium is used as fuel. Mercury heat pipe is used instead of the traditional high temperature heat pipe in order to maintain a low operating temperature of around 350 °C. A superheated steam/water cycle with an efficiency of 33% can ensure the high thermal efficiency of the reactor. The Monte Carlo software MCNP5 is used to analyze the core physics characteristics, including the control rod worth, neutron energy spectrum, neutron distribution, temperature coefficient and core lifetime. A radial power peaking factor of 1.46 with controls inserted can satisfy the heat transfer requirements. A high control rod worth and high transient negative temperature coefficient can also ensure a high inherent safety of this reactor. The core lifetime is around 8 years.
{"title":"Preliminary Design and Neutronics Analysis for a Medium Temperature Heat Pipe Cooled Reactor","authors":"Sina Li, Binhuo Yan, Guojun Tang, Donghong Peng, Hongzhi Lu","doi":"10.1115/icone29-89659","DOIUrl":"https://doi.org/10.1115/icone29-89659","url":null,"abstract":"\u0000 Traditional heat pipe cooled reactors have the characteristics of high core temperature and high fuel enrichment, which makes its development limited by several unfavorable factors, such as the limitation of high temperature materials and low burnup, etc. In this paper, a new heat pipe cooled reactor with a rated thermal power of 1.5MWt is proposed to address the problems of high operating temperature and high fuel enrichment. Low-enriched uranium is used as fuel. Mercury heat pipe is used instead of the traditional high temperature heat pipe in order to maintain a low operating temperature of around 350 °C. A superheated steam/water cycle with an efficiency of 33% can ensure the high thermal efficiency of the reactor. The Monte Carlo software MCNP5 is used to analyze the core physics characteristics, including the control rod worth, neutron energy spectrum, neutron distribution, temperature coefficient and core lifetime. A radial power peaking factor of 1.46 with controls inserted can satisfy the heat transfer requirements. A high control rod worth and high transient negative temperature coefficient can also ensure a high inherent safety of this reactor. The core lifetime is around 8 years.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"32 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":"121151478","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}
Preventing the proliferation of highly enriched uranium (HEU) is an important issue all over the world. The active interrogation of HEU hided in the suitcase based on neutron technique is studied by using MCNP5 code. The neutron source of the detection system is based on the NG-9 neutron generator developed by Northeast Normal University. A set of HEU detection devices has been established. Lead is a kind of neutron-multiplier material with high density, so the lead block is chosen as the substitute for HEU. The size of the aluminum alloy suitcase is 58 cm × 42 cm × 25 cm. The NG-9 D-T neutron generator is placed on the suitcase. A lead block measuring 26.5 cm × 12 cm × 1 cm is placed in the center of the suitcase. The lead block has a density of 11.35 g/cm and a mass of approximately 3.6 kg. Daily clothes are placed inside the suitcase as a distraction. A cylindrical BGO detector with a diameter of 7.2 cm, a height of 29.7 cm is placed close to the suitcase to record gamma rays. A cylindrical lead shield with a thickness of 5 cm is placed outside the BGO detector. Paraffin wax is placed around the whole detection device to protect the neutron radiation and avoid the interference of other substances on the detection results. The purpose of this experiment is to verify the agreement of the MCNP5 simulation results and the experimental results. In this paper, threshold energy neutron analysis (TENA), fast neutron method, and thermal neutron method are used to detect HEU. The simulation results show that the presence of HEU in the suitcase can be determined by neutron flux which is higher than those in the absence of HEU.
{"title":"Active Interrogation of Highly Enriched Uranium in the Suitcase by Using NG-9 Neutron Generator","authors":"Guang Shi, Shiwei Jing","doi":"10.1115/icone29-89597","DOIUrl":"https://doi.org/10.1115/icone29-89597","url":null,"abstract":"\u0000 Preventing the proliferation of highly enriched uranium (HEU) is an important issue all over the world. The active interrogation of HEU hided in the suitcase based on neutron technique is studied by using MCNP5 code. The neutron source of the detection system is based on the NG-9 neutron generator developed by Northeast Normal University. A set of HEU detection devices has been established. Lead is a kind of neutron-multiplier material with high density, so the lead block is chosen as the substitute for HEU. The size of the aluminum alloy suitcase is 58 cm × 42 cm × 25 cm. The NG-9 D-T neutron generator is placed on the suitcase. A lead block measuring 26.5 cm × 12 cm × 1 cm is placed in the center of the suitcase. The lead block has a density of 11.35 g/cm and a mass of approximately 3.6 kg. Daily clothes are placed inside the suitcase as a distraction. A cylindrical BGO detector with a diameter of 7.2 cm, a height of 29.7 cm is placed close to the suitcase to record gamma rays. A cylindrical lead shield with a thickness of 5 cm is placed outside the BGO detector. Paraffin wax is placed around the whole detection device to protect the neutron radiation and avoid the interference of other substances on the detection results. The purpose of this experiment is to verify the agreement of the MCNP5 simulation results and the experimental results. In this paper, threshold energy neutron analysis (TENA), fast neutron method, and thermal neutron method are used to detect HEU. The simulation results show that the presence of HEU in the suitcase can be determined by neutron flux which is higher than those in the absence of HEU.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"143 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":"116191422","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}
Risako Nakashima, Akari Koike, T. Sakai, Norihiro Doda, Masaaki Tanaka
In general, the ultimate heat sink of a decay heat removal system is the atmosphere for accidental conditions of sodium-cooled fast reactor (SFR) designs in Japan. Therefore, risk assessment of external hazards from the atmosphere is important for SFR. However, along with global warming (GW), the number of significant meteorological disasters has increased. Therefore, it is necessary to focus on meteorological phenomena. Hazard curves were developed to address the possibility of abnormal snowfall. By using the hazard curve data, plant dynamics analyses by continuous Markov-chain Monte-Carlo (CMMC) method were conducted with the assumption of failure due to snowfall. The excess frequency of the coolant temperature was quantitatively examined for the design limitation temperature value. The abnormal snowfall height was evaluated using meteorological data. Assuming the possibility of snowfall increase occurring by 2050, the expected value to reproduce 10,000 years was evaluated, and the plant dynamics analyses were conducted using the snowfall height. In conclusion, abnormal snowfall was found to likely increase by GW. In addition, the CMMC method was implemented using the expected value to reproduce 10,000 years, and its effect on the probability of exceeding the temperature limit as the core damage factor was evaluated. As a result, the probability of exceeding this limit increased when GW was considered.
{"title":"Quantitative Risk Assessment With CMMC Method on Abnormal Snowfall Incident for a Sodium-Cooled Fast Reactor","authors":"Risako Nakashima, Akari Koike, T. Sakai, Norihiro Doda, Masaaki Tanaka","doi":"10.1115/icone29-93039","DOIUrl":"https://doi.org/10.1115/icone29-93039","url":null,"abstract":"\u0000 In general, the ultimate heat sink of a decay heat removal system is the atmosphere for accidental conditions of sodium-cooled fast reactor (SFR) designs in Japan. Therefore, risk assessment of external hazards from the atmosphere is important for SFR. However, along with global warming (GW), the number of significant meteorological disasters has increased. Therefore, it is necessary to focus on meteorological phenomena. Hazard curves were developed to address the possibility of abnormal snowfall. By using the hazard curve data, plant dynamics analyses by continuous Markov-chain Monte-Carlo (CMMC) method were conducted with the assumption of failure due to snowfall. The excess frequency of the coolant temperature was quantitatively examined for the design limitation temperature value. The abnormal snowfall height was evaluated using meteorological data. Assuming the possibility of snowfall increase occurring by 2050, the expected value to reproduce 10,000 years was evaluated, and the plant dynamics analyses were conducted using the snowfall height. In conclusion, abnormal snowfall was found to likely increase by GW. In addition, the CMMC method was implemented using the expected value to reproduce 10,000 years, and its effect on the probability of exceeding the temperature limit as the core damage factor was evaluated. As a result, the probability of exceeding this limit increased when GW was considered.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126646585","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}
Pierrick Agullo, A. Gossard, G. Ranchoux, Fabrice Petitot, E. Porcheron
Several nuclear facilities are currently dismantling in France, namely on CEA’s and EDF’s sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling. Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Particularly, we noticed the consequences of roughness on these latter.
{"title":"Relevant Parameters Influencing the Labile Contamination and the Removal Factor","authors":"Pierrick Agullo, A. Gossard, G. Ranchoux, Fabrice Petitot, E. Porcheron","doi":"10.1115/icone29-94236","DOIUrl":"https://doi.org/10.1115/icone29-94236","url":null,"abstract":"\u0000 Several nuclear facilities are currently dismantling in France, namely on CEA’s and EDF’s sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling.\u0000 Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Particularly, we noticed the consequences of roughness on these latter.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"36 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":"125767765","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}
Long Term Operation (LTO) of nuclear power plants (NPPs) will play a key role to reach net zero target. Monitoring and predictive approach detecting in advance faulty SCCs conditions may provide a further key tool in LTO framework. Detecting anomalies may allow the transition from time-based to condition-based predictive maintenance of the NNPs. Predictive algorithms could reduce the number of unplanned outages caused by reactor system failures (one-day outage of a 1000-MW NPPs causes losses of about 500 k$), improving the capacity factor, and keeping high safety margin level of NPPs. To this end, innovative approach by unsupervised machine learning technique (ML) is proposed to detect anomalies of SSCs. Based on principal component analysis and mahalanobis distance is possible to detect in advance the failure of the components. To the purpose a 2D digital twin of primary nuclear pipe under nominal conditions (inner temperature of 300° and an internal pressure of 15.5 MPa) is implemented in finite element code to provide a dataset for unsupervised ML code. The algorithm is then tested under anomaly pattern that deviate from nominal conditions. The results show good code prediction capabilities anticipating the pipe failure. Traditional monitoring combined with ML technique may support LTO program increasing the safety and competitiveness of NPPs.
{"title":"Anomalies Detection in Structures, System and Components for Supporting Nuclear Long Term Operation Program","authors":"S. A. Cancemi, R. Lo Frano","doi":"10.1115/icone29-93193","DOIUrl":"https://doi.org/10.1115/icone29-93193","url":null,"abstract":"\u0000 Long Term Operation (LTO) of nuclear power plants (NPPs) will play a key role to reach net zero target. Monitoring and predictive approach detecting in advance faulty SCCs conditions may provide a further key tool in LTO framework. Detecting anomalies may allow the transition from time-based to condition-based predictive maintenance of the NNPs. Predictive algorithms could reduce the number of unplanned outages caused by reactor system failures (one-day outage of a 1000-MW NPPs causes losses of about 500 k$), improving the capacity factor, and keeping high safety margin level of NPPs. To this end, innovative approach by unsupervised machine learning technique (ML) is proposed to detect anomalies of SSCs. Based on principal component analysis and mahalanobis distance is possible to detect in advance the failure of the components. To the purpose a 2D digital twin of primary nuclear pipe under nominal conditions (inner temperature of 300° and an internal pressure of 15.5 MPa) is implemented in finite element code to provide a dataset for unsupervised ML code. The algorithm is then tested under anomaly pattern that deviate from nominal conditions. The results show good code prediction capabilities anticipating the pipe failure. Traditional monitoring combined with ML technique may support LTO program increasing the safety and competitiveness of NPPs.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"8 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":"127142540","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}
Jiahong Zhu, Jiming Wen, Bo Wang, Yuansheng Lin, Ruifeng Tian
The droplets entrained by the wet steam working medium of nuclear power plant pose a threat to the economy and safety of nuclear power plant. The humidity of steam is directly related to the working efficiency of nuclear power plant. In this paper, the ultrasonic attenuation coefficient in single-phase saturated steam is calculated. Based on the simplified ECAH model in the long wave region, the ultrasonic attenuation coefficient at different frequencies and the relationship between droplet size and humidity in the steam system of nuclear power plant are numerically simulated. For the relationship between droplet size, ultrasonic frequency, steam humidity and ultrasonic energy attenuation coefficient, the results are calculated and analyzed. Results and conclusion: through numerical simulation, the influence of hysteresis attenuation coefficient on the total ultrasonic attenuation coefficient in single-phase saturated steam medium can reach 90%. In single-phase steam medium, the fluctuation of temperature and pressure has little influence on the ultrasonic attenuation coefficient, while the deviation of ultrasonic frequency will have a large error on the ultrasonic attenuation coefficient in steam. The ultrasonic attenuation method is feasible for on-line measurement of steam humidity in nuclear power plant. Compared with the steam environment under other working conditions of nuclear power plant, the ultrasonic attenuation method is more suitable for the measurement of wet steam humidity under the conditions of humidity less than 9% and droplet size greater than 4μm.
{"title":"Numerical Calculation of Ultrasonic Signal Propagation Characteristics in Steam System of Nuclear Power Plant","authors":"Jiahong Zhu, Jiming Wen, Bo Wang, Yuansheng Lin, Ruifeng Tian","doi":"10.1115/icone29-92107","DOIUrl":"https://doi.org/10.1115/icone29-92107","url":null,"abstract":"\u0000 The droplets entrained by the wet steam working medium of nuclear power plant pose a threat to the economy and safety of nuclear power plant. The humidity of steam is directly related to the working efficiency of nuclear power plant. In this paper, the ultrasonic attenuation coefficient in single-phase saturated steam is calculated. Based on the simplified ECAH model in the long wave region, the ultrasonic attenuation coefficient at different frequencies and the relationship between droplet size and humidity in the steam system of nuclear power plant are numerically simulated. For the relationship between droplet size, ultrasonic frequency, steam humidity and ultrasonic energy attenuation coefficient, the results are calculated and analyzed. Results and conclusion: through numerical simulation, the influence of hysteresis attenuation coefficient on the total ultrasonic attenuation coefficient in single-phase saturated steam medium can reach 90%. In single-phase steam medium, the fluctuation of temperature and pressure has little influence on the ultrasonic attenuation coefficient, while the deviation of ultrasonic frequency will have a large error on the ultrasonic attenuation coefficient in steam. The ultrasonic attenuation method is feasible for on-line measurement of steam humidity in nuclear power plant. Compared with the steam environment under other working conditions of nuclear power plant, the ultrasonic attenuation method is more suitable for the measurement of wet steam humidity under the conditions of humidity less than 9% and droplet size greater than 4μm.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"80 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":"131927736","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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