N. Hirota, A. Terada, Xing L. Yan, Kohei Tanaka, Akihito Otani
A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.
{"title":"A Concept of Intermediate Heat Exchanger for High-Temperature Gas Reactor Hydrogen and Power Cogeneration System","authors":"N. Hirota, A. Terada, Xing L. Yan, Kohei Tanaka, Akihito Otani","doi":"10.1115/ICONE26-81718","DOIUrl":"https://doi.org/10.1115/ICONE26-81718","url":null,"abstract":"A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.","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":"86826716","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 problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.
{"title":"Calculational-Experimental Monitoring of Radiation Damage Parameters on VVER Equipment and Their Application During Equipment Residual Life-Time Estimation","authors":"P. Borodkin, Azamat Gazetdinov, N. Khrennikov","doi":"10.1115/ICONE26-81708","DOIUrl":"https://doi.org/10.1115/ICONE26-81708","url":null,"abstract":"The problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.","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":"88283088","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}
Gang Zhao, Xiaoyong Yang, Ping Ye, Jie Wang, W. Peng
High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM), which is designed by Tsinghua university of China, is under construction in Shidao Bay of China. It will be the world’s first pebble-bed type modular HTGR commercial demonstration plant. In HTR-PM project, steam-Rankine cycle has used in the power conversion system because it represents current state-of-the-art technology. Meanwhile, helium turbine for HTGR has been investigated for many years in Tsinghua University. Mock-up machine for HTR-10GT has been built. Helium turbine for 250MW HTGR, which is based on HTR-PM, has completed conceptual designed. However, supercritical carbon dioxide (S-CO2) Brayton cycle has shown to have great potentials for future HTGR technology in recent years because of its critical properties. Helium turbine cycle and S-CO2 Brayton cycle are both candidates for future HTGR. Therefore, comparative study is conducted in this paper. Comparison is focused on achievable efficiencies for each cycle mentioned above and on cycle layout with respect to simplicity and compactness, which primarily determines capital cost. Firstly, the physical model for helium turbine cycle with recuperator and intercooler is built and cycle performance is analyzed based on the parameters of HTR-PM. Then the model for S-CO2 Brayton cycle with recompression is also built and the cycle efficiency is analyzed with the same parameters of HTR-PM. Secondly, comparison between helium turbine cycle and S-CO2 Brayton cycle is made from the view of thermodynamics. Moreover, parameters optimization of both cycles based on HTR-PM is carried out. At last, advantage and drawback of both cycles are discussed from the engineering point. In conclusion, cycle simplicity and technology maturity of helium turbine cycle are better than S-CO2 Brayton cycle. But on the other side, smaller size equipment and less compression work of S-CO2 Brayton Cycle are more competitive than helium turbine cycle. Helium turbine with higher temperature and S-CO2 Brayton Cycle with higher pressure can achieve higher efficiency than steam Rankine cycle.
{"title":"Comparative Study of Helium Turbine Brayton Cycle and Supercritical CO2 Brayton Cycle for HTGR","authors":"Gang Zhao, Xiaoyong Yang, Ping Ye, Jie Wang, W. Peng","doi":"10.1115/ICONE26-81561","DOIUrl":"https://doi.org/10.1115/ICONE26-81561","url":null,"abstract":"High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM), which is designed by Tsinghua university of China, is under construction in Shidao Bay of China. It will be the world’s first pebble-bed type modular HTGR commercial demonstration plant. In HTR-PM project, steam-Rankine cycle has used in the power conversion system because it represents current state-of-the-art technology. Meanwhile, helium turbine for HTGR has been investigated for many years in Tsinghua University. Mock-up machine for HTR-10GT has been built. Helium turbine for 250MW HTGR, which is based on HTR-PM, has completed conceptual designed. However, supercritical carbon dioxide (S-CO2) Brayton cycle has shown to have great potentials for future HTGR technology in recent years because of its critical properties. Helium turbine cycle and S-CO2 Brayton cycle are both candidates for future HTGR. Therefore, comparative study is conducted in this paper. Comparison is focused on achievable efficiencies for each cycle mentioned above and on cycle layout with respect to simplicity and compactness, which primarily determines capital cost. Firstly, the physical model for helium turbine cycle with recuperator and intercooler is built and cycle performance is analyzed based on the parameters of HTR-PM. Then the model for S-CO2 Brayton cycle with recompression is also built and the cycle efficiency is analyzed with the same parameters of HTR-PM. Secondly, comparison between helium turbine cycle and S-CO2 Brayton cycle is made from the view of thermodynamics. Moreover, parameters optimization of both cycles based on HTR-PM is carried out. At last, advantage and drawback of both cycles are discussed from the engineering point. In conclusion, cycle simplicity and technology maturity of helium turbine cycle are better than S-CO2 Brayton cycle. But on the other side, smaller size equipment and less compression work of S-CO2 Brayton Cycle are more competitive than helium turbine cycle. Helium turbine with higher temperature and S-CO2 Brayton Cycle with higher pressure can achieve higher efficiency than steam Rankine cycle.","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":"88708077","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 order to guarantee NPP operation safety, it is necessary to ensure the reliability operation of safety related equipment by equipment qualification. In the past, plant equipment qualification function requirement is an envelope requirement based on engineering experience, and it is not beneficial to NPP economical efficiency. These days, representative Gen III NPP (e.g. AP1000 and EPR) adopt improved technology in equipment qualification function requirement design, and more accurate requirement is designed. In this paper, AP1000 and EPR equipment qualification function requirement design methodology is studied and analyzed as the first step. Then, a safety related equipment qualification function requirement design methodology which is applicable for China self intellectual property Gen III NPP is provided. Furthermore, an example of equipment qualification function requirement design is carried out by analyzing nuclear instrument system power range channel sensor.
{"title":"Study on Advanced PWR NPP Safety Related Equipment Qualification Function Requirement Design Methodology","authors":"Xu Zhao, Miao Zhuang, Yi Ke","doi":"10.1115/ICONE26-81212","DOIUrl":"https://doi.org/10.1115/ICONE26-81212","url":null,"abstract":"In order to guarantee NPP operation safety, it is necessary to ensure the reliability operation of safety related equipment by equipment qualification. In the past, plant equipment qualification function requirement is an envelope requirement based on engineering experience, and it is not beneficial to NPP economical efficiency. These days, representative Gen III NPP (e.g. AP1000 and EPR) adopt improved technology in equipment qualification function requirement design, and more accurate requirement is designed. In this paper, AP1000 and EPR equipment qualification function requirement design methodology is studied and analyzed as the first step. Then, a safety related equipment qualification function requirement design methodology which is applicable for China self intellectual property Gen III NPP is provided. Furthermore, an example of equipment qualification function requirement design is carried out by analyzing nuclear instrument system power range channel sensor.","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":"79056906","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}
Mo Tao, Ruotong Qu, Zhiwu Ke, Zhaoxu Chen, Xianling Li, Yi Feng
Steam generator (SG) is one of the key equipment of nuclear power units. Because of the large range of its loads changing, the water level control of SG effectively is an essential secure guarantee of nuclear power plants. SG is a complex system, besides imbalance and non-minimum phase characteristic, it also has the properties of nonlinearity, time-varying and with small stability margin. There are many difficulties in water level control of SG. Of which false water level and varying parameters are the most severe problems. In this paper, first the water level features and the water level control principle of U-tube steam generator (UTSG) are introduced. Then mathematical model mechanism and both the static and dynamic characteristic of it water level are discussed. Finally various control methods are used for comparing the control effect. Intelligent control is a type of control strategy which imitates human intelligence behavior. It is mainly aimed at the controlled plant with complicate model parameters, or which model structure hard to describe accurately by mathematical method. Cloud Model theory is proposed by Academician Li Deyi based on the idea of artificial intelligence with uncertainty. This theory focus on analyzing the uncertainty of control plant, realizes the uncertain conversion between qualitative concept and quantitative numerical by combining ambiguity and randomness. In the field of control technology, ambiguity and randomness make it difficult to establishing precise mathematical model of control plant, and become a bottleneck during the research of improving stability, accuracy and quickness of control system. In this context, Cloud Model can be a good conversion between qualitative concept and quantitative numerical due to its ability of showing the uncertainty of qualitative concept which described by natural language. Under the action of external input, system control can be realized by inferencing according to the qualitative concept and uncertainty rules of Cloud Model. In this paper, the researched Cloud Model control system is based on normal distribution, because a large number of random events in nature and society obey or approximately obey normal distribution. The rate of convergence of Cloud Model control is evidently faster than PID. Moreover, the capability of Cloud Model control in tracking, adapting, anti-interference and overcoming large time lag are apparently superior when comparing with the control effect of PID.
{"title":"The Cloud Model Theory of Intelligent Control Method for Non-Minimum-Phase and Non-Self-Balancing System in Nuclear Power","authors":"Mo Tao, Ruotong Qu, Zhiwu Ke, Zhaoxu Chen, Xianling Li, Yi Feng","doi":"10.1115/ICONE26-81829","DOIUrl":"https://doi.org/10.1115/ICONE26-81829","url":null,"abstract":"Steam generator (SG) is one of the key equipment of nuclear power units. Because of the large range of its loads changing, the water level control of SG effectively is an essential secure guarantee of nuclear power plants. SG is a complex system, besides imbalance and non-minimum phase characteristic, it also has the properties of nonlinearity, time-varying and with small stability margin. There are many difficulties in water level control of SG. Of which false water level and varying parameters are the most severe problems.\u0000 In this paper, first the water level features and the water level control principle of U-tube steam generator (UTSG) are introduced. Then mathematical model mechanism and both the static and dynamic characteristic of it water level are discussed. Finally various control methods are used for comparing the control effect.\u0000 Intelligent control is a type of control strategy which imitates human intelligence behavior. It is mainly aimed at the controlled plant with complicate model parameters, or which model structure hard to describe accurately by mathematical method. Cloud Model theory is proposed by Academician Li Deyi based on the idea of artificial intelligence with uncertainty. This theory focus on analyzing the uncertainty of control plant, realizes the uncertain conversion between qualitative concept and quantitative numerical by combining ambiguity and randomness. In the field of control technology, ambiguity and randomness make it difficult to establishing precise mathematical model of control plant, and become a bottleneck during the research of improving stability, accuracy and quickness of control system.\u0000 In this context, Cloud Model can be a good conversion between qualitative concept and quantitative numerical due to its ability of showing the uncertainty of qualitative concept which described by natural language. Under the action of external input, system control can be realized by inferencing according to the qualitative concept and uncertainty rules of Cloud Model. In this paper, the researched Cloud Model control system is based on normal distribution, because a large number of random events in nature and society obey or approximately obey normal distribution.\u0000 The rate of convergence of Cloud Model control is evidently faster than PID. Moreover, the capability of Cloud Model control in tracking, adapting, anti-interference and overcoming large time lag are apparently superior when comparing with the control effect of PID.","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":"86427943","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}
E. Babeshko, V. Kharchenko, Kostiantyn Leontiiev, O. Odarushchenko, Oleksiy Strjuk
Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.
{"title":"NPP I&C Safety Assessment by Aggregation of Formal Techniques","authors":"E. Babeshko, V. Kharchenko, Kostiantyn Leontiiev, O. Odarushchenko, Oleksiy Strjuk","doi":"10.1115/ICONE26-82270","DOIUrl":"https://doi.org/10.1115/ICONE26-82270","url":null,"abstract":"Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.","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":"87623325","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
With the development of small modular reactors, the hydrogen risk reducing technology cannot be ignored. Special safety facilities of small modular reactor (SMR) are investigated and studied, and a serious accident analysis program model for SMR is established. The combination of Pre-inerting and hydrogen recombination was used to control the hydrogen risk. The effectiveness of the hydrogen control system is analyzed by using the GASFLOW program. The results show that the volume fraction of hydrogen in the containment dome is higher than that in the other parts of the containment during the calculation. Because of the small size and tight internal structure, hydrogen accumulates in the narrow channel, which increases the hydrogen concentration in the local channel. Inerting reduces the concentration of oxygen in the containment and effectively controls the possibility of flame acceleration and blasting transition in high hydrogen concentration regions.
{"title":"Hydrogen Risk Reducing Technology in Small Modular Reactor","authors":"Yanlin Chen, Xuefeng Lv","doi":"10.1115/ICONE26-81705","DOIUrl":"https://doi.org/10.1115/ICONE26-81705","url":null,"abstract":"With the development of small modular reactors, the hydrogen risk reducing technology cannot be ignored. Special safety facilities of small modular reactor (SMR) are investigated and studied, and a serious accident analysis program model for SMR is established. The combination of Pre-inerting and hydrogen recombination was used to control the hydrogen risk. The effectiveness of the hydrogen control system is analyzed by using the GASFLOW program. The results show that the volume fraction of hydrogen in the containment dome is higher than that in the other parts of the containment during the calculation. Because of the small size and tight internal structure, hydrogen accumulates in the narrow channel, which increases the hydrogen concentration in the local channel. Inerting reduces the concentration of oxygen in the containment and effectively controls the possibility of flame acceleration and blasting transition in high hydrogen concentration regions.","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":"90676918","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}
From the general industrial control system to the nuclear power plant control platform, the threat of information security has its own particularity more than continuity. The original dedicated system in general industrial area is gradually replaced by many common protocol, software and equipment. As a result, the security vulnerabilities are more likely to be used illegally. For a specific nuclear power plant digital control platform-NASPIC, the vulnerability analysis of platform is performed. Mainly two aspects of technology and management are to be carried out. For technical aspects, four categories problems-unauthorized execution, unauthorized write, unauthorized reading and reject service-are analyzed. Management is mainly about the lack of management strategy and strategy vulnerability. By analyzing the fragility of the instrument control platform, the key equipments, key channels and key modules are proposed. The qualitative and quantitative rules are deduced for evaluation of NASPIC information security.
{"title":"The Security Vulnerability Analysis of Nuclear Power Digital Instrument Control Platform NASPIC","authors":"Hua Liu, Xiaohua Yang, Zhaohui Liu, Meng Li, Zhi Chen, Zhigang Feng, Qingya Zhao","doi":"10.1115/ICONE26-81486","DOIUrl":"https://doi.org/10.1115/ICONE26-81486","url":null,"abstract":"From the general industrial control system to the nuclear power plant control platform, the threat of information security has its own particularity more than continuity. The original dedicated system in general industrial area is gradually replaced by many common protocol, software and equipment. As a result, the security vulnerabilities are more likely to be used illegally. For a specific nuclear power plant digital control platform-NASPIC, the vulnerability analysis of platform is performed. Mainly two aspects of technology and management are to be carried out. For technical aspects, four categories problems-unauthorized execution, unauthorized write, unauthorized reading and reject service-are analyzed. Management is mainly about the lack of management strategy and strategy vulnerability. By analyzing the fragility of the instrument control platform, the key equipments, key channels and key modules are proposed. The qualitative and quantitative rules are deduced for evaluation of NASPIC information security.","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":"80678322","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nuclear safety is one of the key issues for a nuclear power plant (NPP). The alarm system plays a critical role for the safe and efficient operation of an NPP which affects the correctness and efficiency of the operators in dealing with the accidents. It is even more important for the alarm system of a multi-modular NPP which has more than one reactor modules in a single unit because all the modules are usually monitored in the same main control room. The alarm generation and display mechanism of a typical multi-modular NPP, the High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM), is analyzed in this paper which has two reactor modules coupled to one steam turbine. Three operators are responsible for the operation of two nuclear islands and a conventional island, respectively. The alarm generation and display processes will be discussed in this paper. Firstly, the architecture of the RPS and the alarm system related to the red and yellow alarms is introduced. Then the generation and display mechanism of the red and yellow alarms is proposed. A protection variable of a design basis accident is given as an example for the alarm signal handling. The characteristics of the alarm system are then discussed. More optimization directions on the alarm design for multi-modular NPPs are proposed in the end.
{"title":"Analysis of the Alarm Generation and Display for the Reactor Accidents in HTR-PM","authors":"Chao Guo, Shuqiao Zhou, Duo Li, Xiaojin Huang","doi":"10.1115/ICONE26-82483","DOIUrl":"https://doi.org/10.1115/ICONE26-82483","url":null,"abstract":"Nuclear safety is one of the key issues for a nuclear power plant (NPP). The alarm system plays a critical role for the safe and efficient operation of an NPP which affects the correctness and efficiency of the operators in dealing with the accidents. It is even more important for the alarm system of a multi-modular NPP which has more than one reactor modules in a single unit because all the modules are usually monitored in the same main control room. The alarm generation and display mechanism of a typical multi-modular NPP, the High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM), is analyzed in this paper which has two reactor modules coupled to one steam turbine. Three operators are responsible for the operation of two nuclear islands and a conventional island, respectively. The alarm generation and display processes will be discussed in this paper. Firstly, the architecture of the RPS and the alarm system related to the red and yellow alarms is introduced. Then the generation and display mechanism of the red and yellow alarms is proposed. A protection variable of a design basis accident is given as an example for the alarm signal handling. The characteristics of the alarm system are then discussed. More optimization directions on the alarm design for multi-modular NPPs are proposed in the end.","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":"82029977","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nuclear power plants typically consider a turbine trip and rapid closure of the main turbine stop valves as a normal transient event. As required by ASME Code [1], the piping loads generated by the unbalanced pressures in the system resulting from the rapid valve closure are part of the analyzed spectrum of conditions in the piping and support analysis. The analysis that determines the magnitude and timing of the loads is often referred to as a “steamhammer” analysis. Currently, there are several computerized analytical techniques to determine the steamhammer piping and support loads [2], but because of compressibility assumptions the equations become more difficult to solve than in the analogous incompressible waterhammer models, which are quite straightforward. This paper highlights the effect of fluid compressibility by comparing results predicted by both waterhammer (slightly compressible) flow models and compressible (steamhammer) flow models. Guidelines are offered to show how parameters of a piping system (such as pipe length, valve closure time and flow characteristic, steam initial state properties, and velocity) can be interpreted to determine if compressible effects are insignificant or if they play a significant role.
{"title":"The Effects of Compressibility and Piping Geometry on Steamhammer Loads","authors":"F. Moody, Robert J. Stakenborghs","doi":"10.1115/ICONE26-81003","DOIUrl":"https://doi.org/10.1115/ICONE26-81003","url":null,"abstract":"Nuclear power plants typically consider a turbine trip and rapid closure of the main turbine stop valves as a normal transient event. As required by ASME Code [1], the piping loads generated by the unbalanced pressures in the system resulting from the rapid valve closure are part of the analyzed spectrum of conditions in the piping and support analysis. The analysis that determines the magnitude and timing of the loads is often referred to as a “steamhammer” analysis.\u0000 Currently, there are several computerized analytical techniques to determine the steamhammer piping and support loads [2], but because of compressibility assumptions the equations become more difficult to solve than in the analogous incompressible waterhammer models, which are quite straightforward.\u0000 This paper highlights the effect of fluid compressibility by comparing results predicted by both waterhammer (slightly compressible) flow models and compressible (steamhammer) flow models. Guidelines are offered to show how parameters of a piping system (such as pipe length, valve closure time and flow characteristic, steam initial state properties, and velocity) can be interpreted to determine if compressible effects are insignificant or if they play a significant role.","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":"78214770","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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