L. Ying, Wang Ya-feng, Pan Bo, Tang Lei, Feng Bo, Cao Guo-hai
With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I&C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make basement for evaluating the reliability and safety of DCS.
{"title":"Prediction and Sensibility Analysis for Nuclear Safety-Critical Software Reliability of DCS","authors":"L. Ying, Wang Ya-feng, Pan Bo, Tang Lei, Feng Bo, Cao Guo-hai","doi":"10.1115/ICONE26-81647","DOIUrl":"https://doi.org/10.1115/ICONE26-81647","url":null,"abstract":"With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I&C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make basement for evaluating the reliability and safety of DCS.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73908460","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}
A novel combined small modular reactor (SMR) and gas turbine cycle is presented. This SMR-GT cycle is modeled using fundamental thermodynamic relationships and compared to existing state-of-the-art power generation cycles. The SMR-GT cycle includes an 82 MWe SMR cycle that is combined with a 54 MWe gas turbine cycle. A heat exchanger is used to extract energy from the gas turbine exhaust to create superheated main steam and provide reheat downstream of the LP turbine. This results in a 32 MWe increase in the SMR cycle for total unit output of 136 MWe. Comparisons of thermal efficiency, heat rate, CO2 emissions, and net generation are made between a stand-alone SMR, a typical combined cycle gas turbine (CCGT), standalone gas turbine and the combined SMR-GT cycles. Several advantages of the SMR-GT cycle are discussed. In addition, the rapid deployment of a gas turbine allows for a power station to deliver power and earn revenue prior to completion of the more complex SMR portion of the plant. The SMR portion of the cycle also reduces the overall fuel cost volatility associated with gas turbine based power station.
{"title":"A Combined Small Modular Reactor and Gas Turbine Cycle With Reheat","authors":"Robert J. Stakenborghs, G. Kramer","doi":"10.1115/ICONE26-81002","DOIUrl":"https://doi.org/10.1115/ICONE26-81002","url":null,"abstract":"A novel combined small modular reactor (SMR) and gas turbine cycle is presented. This SMR-GT cycle is modeled using fundamental thermodynamic relationships and compared to existing state-of-the-art power generation cycles. The SMR-GT cycle includes an 82 MWe SMR cycle that is combined with a 54 MWe gas turbine cycle. A heat exchanger is used to extract energy from the gas turbine exhaust to create superheated main steam and provide reheat downstream of the LP turbine. This results in a 32 MWe increase in the SMR cycle for total unit output of 136 MWe.\u0000 Comparisons of thermal efficiency, heat rate, CO2 emissions, and net generation are made between a stand-alone SMR, a typical combined cycle gas turbine (CCGT), standalone gas turbine and the combined SMR-GT cycles. Several advantages of the SMR-GT cycle are discussed.\u0000 In addition, the rapid deployment of a gas turbine allows for a power station to deliver power and earn revenue prior to completion of the more complex SMR portion of the plant. The SMR portion of the cycle also reduces the overall fuel cost volatility associated with gas turbine based power station.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90607938","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}
V. Kharchenko, Andriy Kovalenko, Kostiantyn Leontiiev, A. Panarin, Vyacheslav Duzhy
Diversity approach is used to decrease risk of common cause failure (CCF) of Nuclear Power Plant (NPP) Instrumentation and Control systems (I&Cs). Application of a multi-diversity, i.e. a few different types of version redundancy allows minimizing CCF risk. On the other side, implementation of diversity increases cost and complicates maintenance of multi-version I&Cs. Hence, it is important to find optimal solution according with criteria “required level of diversity (safety) / minimal cost and maintenance complexity. Modern FPGA technology creates additional possibilities to meet requirements of the standards (such as NUREG/CR-7007, IEEE Std 7-4.3.2-2016, IAEA SSR-2/1:2016, IAEA NP-T-3.17:2016 and others) by developing main and diverse subsystems on the basis of the same FPGA platform. Existing diversity normative base should be enhanced in three directions — scope, depth and rigor to provide more detailed description of possible applied techniques and tools for quantitative assessment. The goals of the paper which overviews practical issues of diversity application are the following: - present extended classification of diversity considering additional types of version redundancy for FPGA platform based I&Cs (logical processing equipment, life cycle, logic/algorithm etc.) in comparing to NUREG7007; - describe the modified technique of diversity assessment taking into account three and more levels of diversity classification; - illustrate and discuss variants of assurance of the required degree of diversity by use of the RadICS FPGA platform to develop main and diverse subsystems. The classification is specified considering diversity of hardware and FPGA designs. In particular, diversity of hard logic and soft processors, interfaces and buses, self-diagnostics means and others are described and embedded into NUREG/CR-7007 classification. The NUREG7007-based diversity assessment techniques supporting all stage of analyzing options are discussed, and algorithms for versions choice are described. This technique takes into account more detailed specification of diversity classification (for types, subtypes and sub-subtypes of diversity for logic diversity, logic processing equipment diversity and others) and options to evaluate weight coefficients. Case study is based on description of two options of RadICS FPGA platform application to develop two-version NPP I&C, which meets standard requirements to diversity.
{"title":"Multi-Diversity for FPGA Platform Based NPP I&C Systems: New Possibilities and Assessment Technique","authors":"V. Kharchenko, Andriy Kovalenko, Kostiantyn Leontiiev, A. Panarin, Vyacheslav Duzhy","doi":"10.1115/ICONE26-82377","DOIUrl":"https://doi.org/10.1115/ICONE26-82377","url":null,"abstract":"Diversity approach is used to decrease risk of common cause failure (CCF) of Nuclear Power Plant (NPP) Instrumentation and Control systems (I&Cs). Application of a multi-diversity, i.e. a few different types of version redundancy allows minimizing CCF risk. On the other side, implementation of diversity increases cost and complicates maintenance of multi-version I&Cs. Hence, it is important to find optimal solution according with criteria “required level of diversity (safety) / minimal cost and maintenance complexity. Modern FPGA technology creates additional possibilities to meet requirements of the standards (such as NUREG/CR-7007, IEEE Std 7-4.3.2-2016, IAEA SSR-2/1:2016, IAEA NP-T-3.17:2016 and others) by developing main and diverse subsystems on the basis of the same FPGA platform. Existing diversity normative base should be enhanced in three directions — scope, depth and rigor to provide more detailed description of possible applied techniques and tools for quantitative assessment.\u0000 The goals of the paper which overviews practical issues of diversity application are the following:\u0000 - present extended classification of diversity considering additional types of version redundancy for FPGA platform based I&Cs (logical processing equipment, life cycle, logic/algorithm etc.) in comparing to NUREG7007;\u0000 - describe the modified technique of diversity assessment taking into account three and more levels of diversity classification;\u0000 - illustrate and discuss variants of assurance of the required degree of diversity by use of the RadICS FPGA platform to develop main and diverse subsystems.\u0000 The classification is specified considering diversity of hardware and FPGA designs. In particular, diversity of hard logic and soft processors, interfaces and buses, self-diagnostics means and others are described and embedded into NUREG/CR-7007 classification.\u0000 The NUREG7007-based diversity assessment techniques supporting all stage of analyzing options are discussed, and algorithms for versions choice are described. This technique takes into account more detailed specification of diversity classification (for types, subtypes and sub-subtypes of diversity for logic diversity, logic processing equipment diversity and others) and options to evaluate weight coefficients.\u0000 Case study is based on description of two options of RadICS FPGA platform application to develop two-version NPP I&C, which meets standard requirements to diversity.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89570104","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}
Yi Feng, Xianling Li, Zhiwu Ke, Zhaoxu Chen, Mo Tao
In nuclear power plant system, pump is the key equipment to maintain the flow of the primary loop coolant and the secondary loop heat transfer fluid. The main coolant pump and the feed water pump are in long-term operation status. Bearings are the key components to ensure stable operation of the pump, and which could be damaged in abnormal conditions. Once the failure occurred in the bearings, pumps would exhibit periodic vibration, which might cause the flow pulsations of coolant and heat transfer fluid; gradually, these situations could reduce the control accuracy and the stability of pump. Therefore, the detection and diagnosis of pump bearings are significant to improve the safety and stability of reactor system. We proposed an approach combined with signal processing and machine learning to extract the signal features and recognize the signal samples automatically. The proposed approach consists of three main steps: firstly, empirical mode decomposition (EMD) is applied to decompose the signals into several intrinsic mode functions (IMFs) which are corresponding to the different components of the original signals; secondly, calculating the correlation coefficient between each IMF and the original signal, the correlation coefficient sequence imply the components distribution of the signal which can be applied to recognize the signal samples; finally, extracting a part of correlation coefficient sequences to train the support vector machine (SVM), and then an classifier can be obtained and use to recognize the other signal samples automatically. Experimental results show that this method can effectively detect the pump bearing operating conditions and failures, and can provide a reference for the safe and stable operation of reactor pumps.
{"title":"Pump Bearing Fault Detection Based on EMD and SVM","authors":"Yi Feng, Xianling Li, Zhiwu Ke, Zhaoxu Chen, Mo Tao","doi":"10.1115/ICONE26-81584","DOIUrl":"https://doi.org/10.1115/ICONE26-81584","url":null,"abstract":"In nuclear power plant system, pump is the key equipment to maintain the flow of the primary loop coolant and the secondary loop heat transfer fluid. The main coolant pump and the feed water pump are in long-term operation status. Bearings are the key components to ensure stable operation of the pump, and which could be damaged in abnormal conditions. Once the failure occurred in the bearings, pumps would exhibit periodic vibration, which might cause the flow pulsations of coolant and heat transfer fluid; gradually, these situations could reduce the control accuracy and the stability of pump. Therefore, the detection and diagnosis of pump bearings are significant to improve the safety and stability of reactor system. We proposed an approach combined with signal processing and machine learning to extract the signal features and recognize the signal samples automatically. The proposed approach consists of three main steps: firstly, empirical mode decomposition (EMD) is applied to decompose the signals into several intrinsic mode functions (IMFs) which are corresponding to the different components of the original signals; secondly, calculating the correlation coefficient between each IMF and the original signal, the correlation coefficient sequence imply the components distribution of the signal which can be applied to recognize the signal samples; finally, extracting a part of correlation coefficient sequences to train the support vector machine (SVM), and then an classifier can be obtained and use to recognize the other signal samples automatically. Experimental results show that this method can effectively detect the pump bearing operating conditions and failures, and can provide a reference for the safe and stable operation of reactor pumps.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83031361","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}
Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required. Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.
{"title":"A Novel More Reliable and Extensible Architecture of Instrumentation and Control Systems","authors":"Shuqiao Zhou, Chao Guo, Duo Li, Xiaojin Huang","doi":"10.1115/ICONE26-81570","DOIUrl":"https://doi.org/10.1115/ICONE26-81570","url":null,"abstract":"Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required.\u0000 Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89870663","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 sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device. The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect. In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.
{"title":"Geometry Survey on the Convex Shaped Core for Recriticality Prevention Against CDA in Sodium-Cooled Fast Reactor","authors":"K. Chitose, Y. Tachi, T. Wakabayashi, N. Takaki","doi":"10.1115/ICONE26-81331","DOIUrl":"https://doi.org/10.1115/ICONE26-81331","url":null,"abstract":"In sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device.\u0000 The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect.\u0000 In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78097697","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}
Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity. Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.
{"title":"Application of Monte Carlo Methods in Reactor Protection System Reliability Research","authors":"Duo Li, Zhaojun Hao, Shuqiao Zhou, Chao Guo","doi":"10.1115/ICONE26-81300","DOIUrl":"https://doi.org/10.1115/ICONE26-81300","url":null,"abstract":"Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity.\u0000 Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78462318","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}
Quan-hu Zhang, Wenming Zuo, Sufen Li, S. Hou, Lin Zhuang, Wenheng Zhou
Gamma camera imaging technology is a non-destructive passive radiation imaging technology, which can quickly find the unknown source location, search the exact number of radioactive sources and relative intensity. Therefore, it is very important and widely used in the fields of effective regulation of radioactive sources, handling of various nuclear emergencies, nuclear arms control and other fields. In the practical application of gamma camera, it often faces the imaging difference caused by the difference of radiation source intensity, detection time and detection distance. It is helpful to study the change of imaging characteristics under different experimental conditions for the practical application of gamma camera under different scenes. In this paper, the structure and imaging principle of gamma camera are analyzed in detail. Using the Cartogam portable gamma camera, a set of comparative experiments are carried out to study the time characteristics, distance characteristics and source intensity characteristics of the gamma camera. The results show that the imaging quality of gamma camera is positively correlated with the time source intensity, negatively correlated with the distance. For a milliCurie source, the gamma camera has very good fast-position resolution at a distance of 1 meter from the radioactive source and can form a more complete hot spot image within 5 minutes. When the distance becomes larger, the radioactive source needs at least 20 minutes to form a more accurate hot spot image. The hot spot is no longer as complete as a concentric circle structure, but can achieve precise positioning. For a strong source of more than ten milliCurie, immediate imaging within two minutes can be basically achieved within two meters. Under multi-source conditions, when the source intensities differ greatly and the distance between sources is relatively close, the detection of weak source can not be achieved by increasing the measurement time. However, by observing the counting images in a short period of time, the possibility of existence of a weak source can be deduced. Therefore, in the practical application of the gamma camera, it is necessary to constantly adjust its imaging conditions to ensure the detection of weak source verification. In this paper, the Monte Carlo model of gamma camera is set up to simulate the imaging. Compared with the actual imaging hot spots, the simulated images can correctly reflect the hot spot graph’s level distribution, which has the value of further research.
{"title":"Research on Gamma Camera Imaging Characteristics","authors":"Quan-hu Zhang, Wenming Zuo, Sufen Li, S. Hou, Lin Zhuang, Wenheng Zhou","doi":"10.1115/ICONE26-81936","DOIUrl":"https://doi.org/10.1115/ICONE26-81936","url":null,"abstract":"Gamma camera imaging technology is a non-destructive passive radiation imaging technology, which can quickly find the unknown source location, search the exact number of radioactive sources and relative intensity. Therefore, it is very important and widely used in the fields of effective regulation of radioactive sources, handling of various nuclear emergencies, nuclear arms control and other fields. In the practical application of gamma camera, it often faces the imaging difference caused by the difference of radiation source intensity, detection time and detection distance. It is helpful to study the change of imaging characteristics under different experimental conditions for the practical application of gamma camera under different scenes. In this paper, the structure and imaging principle of gamma camera are analyzed in detail. Using the Cartogam portable gamma camera, a set of comparative experiments are carried out to study the time characteristics, distance characteristics and source intensity characteristics of the gamma camera. The results show that the imaging quality of gamma camera is positively correlated with the time source intensity, negatively correlated with the distance. For a milliCurie source, the gamma camera has very good fast-position resolution at a distance of 1 meter from the radioactive source and can form a more complete hot spot image within 5 minutes. When the distance becomes larger, the radioactive source needs at least 20 minutes to form a more accurate hot spot image. The hot spot is no longer as complete as a concentric circle structure, but can achieve precise positioning. For a strong source of more than ten milliCurie, immediate imaging within two minutes can be basically achieved within two meters. Under multi-source conditions, when the source intensities differ greatly and the distance between sources is relatively close, the detection of weak source can not be achieved by increasing the measurement time. However, by observing the counting images in a short period of time, the possibility of existence of a weak source can be deduced. Therefore, in the practical application of the gamma camera, it is necessary to constantly adjust its imaging conditions to ensure the detection of weak source verification. In this paper, the Monte Carlo model of gamma camera is set up to simulate the imaging. Compared with the actual imaging hot spots, the simulated images can correctly reflect the hot spot graph’s level distribution, which has the value of further research.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81238019","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 paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used. The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed. The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.
{"title":"The Influence of Centrifugal Pump Characteristics on Dynamic Loadings on Pipelines After Power Failure","authors":"J. Marcinkiewicz, K. Karaśkiewicz, Claes Joheman","doi":"10.1115/ICONE26-81825","DOIUrl":"https://doi.org/10.1115/ICONE26-81825","url":null,"abstract":"The paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used.\u0000 The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed.\u0000 The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82314549","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}
iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.
{"title":"iB1350: Part 2 — Level1 PRA Considering Optimization of Safety Systems for the iB1350","authors":"T. Go, Sato Takashi, Komori Yuji, M. Keiji","doi":"10.1115/ICONE26-82552","DOIUrl":"https://doi.org/10.1115/ICONE26-82552","url":null,"abstract":"iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76849148","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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