M. Liao, C. Dai, can ma, Yong Liu, Zheng-Xing Zhao, Zhouyang Liu
The gas/liquid metal magnetohydrodynamic generator (G/LM-MHD) with the mixture of gas and liquid metal as working fluids shows a promising future due to recent development of liquid metal cooled nuclear reactors. Previous efforts on the G/LM-MHD energy conversion systems have predicted a higher efficiency than traditional thermodynamics cycle. However, most of the earlier studies focus on the conception designs, feasibility analysis and preliminary experiments, while less attention paid on some specific problems such as the bubble phenomenon in the two-phase flow. Therefore, this paper deals with numerical study on the performance characteristics of the gas/liquid metal two-phase flow in an ideal Faraday-type MHD channel, of which the geometry structure is 30 × 30 × 80 mm cuboid segmentary style. The conductive mixture fluid is composed of nitrogen as the gas phase and gallium as the liquid phase (N2/Ga). The temperature at the channel inlet is about 600 K considering the heat transfer after the mixing chamber, while the inlet velocity is around 10 m/s and gas volumetric void fraction is 50%. The external magnetic field is assumed as 4 Tesla adopting the superconducting technology, which seems essential for MHD industrial practice. Then the simulation is accomplished using a modified two-phase mixture model considering the electromagnetic influence. The simulation results show that the distribution of temperature changes much weaker than pressure and velocity, which agrees with earlier one-dimension analysis. On the other hand, the results characterizes clearly the increase of the void fraction close to the electrodes, which can explain intuitively the decrease of the power-generating capacity. Besides, the power output is predicted to reach maximum 22.5 kW while the voltage is 1.2 V and the power density can be 312.5 MW/m3 which is far beyond traditional steam turbines. This study shows a promising future of the gas/liquid metal MHD generator for the small nuclear plants and power systems.
{"title":"Numerical Study on the Two-Phase Flow for a Gas/Liquid Metal Magnetohydrodynamic Generator","authors":"M. Liao, C. Dai, can ma, Yong Liu, Zheng-Xing Zhao, Zhouyang Liu","doi":"10.1115/ICONE26-82231","DOIUrl":"https://doi.org/10.1115/ICONE26-82231","url":null,"abstract":"The gas/liquid metal magnetohydrodynamic generator (G/LM-MHD) with the mixture of gas and liquid metal as working fluids shows a promising future due to recent development of liquid metal cooled nuclear reactors. Previous efforts on the G/LM-MHD energy conversion systems have predicted a higher efficiency than traditional thermodynamics cycle. However, most of the earlier studies focus on the conception designs, feasibility analysis and preliminary experiments, while less attention paid on some specific problems such as the bubble phenomenon in the two-phase flow. Therefore, this paper deals with numerical study on the performance characteristics of the gas/liquid metal two-phase flow in an ideal Faraday-type MHD channel, of which the geometry structure is 30 × 30 × 80 mm cuboid segmentary style. The conductive mixture fluid is composed of nitrogen as the gas phase and gallium as the liquid phase (N2/Ga). The temperature at the channel inlet is about 600 K considering the heat transfer after the mixing chamber, while the inlet velocity is around 10 m/s and gas volumetric void fraction is 50%. The external magnetic field is assumed as 4 Tesla adopting the superconducting technology, which seems essential for MHD industrial practice. Then the simulation is accomplished using a modified two-phase mixture model considering the electromagnetic influence. The simulation results show that the distribution of temperature changes much weaker than pressure and velocity, which agrees with earlier one-dimension analysis. On the other hand, the results characterizes clearly the increase of the void fraction close to the electrodes, which can explain intuitively the decrease of the power-generating capacity. Besides, the power output is predicted to reach maximum 22.5 kW while the voltage is 1.2 V and the power density can be 312.5 MW/m3 which is far beyond traditional steam turbines. This study shows a promising future of the gas/liquid metal MHD generator for the small nuclear plants and power systems.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"332 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74979926","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}
High breeding with light water cooling has been studied for decades, though is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. To harden the neutron spectrum and thereby to enhance the fuel utilization or even to achieve breeding with light water cooling, the tight-lattice assembly was proposed and applied to High Conversion LWRs. Nonetheless, none of them achieved high breeding. Until recently, the tightly packed fuel assembly (TPFA) is designed for the purpose of high breeding. The ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to be less than 0.1. Super Fast Breeding Reactor (Super FBR) adopts TPFA and achieves breeding performance with compound system doubling time (CSDT) of 43 years. In this study, the breeder BWR core also applies TPFA and achieves CSDT of 50 years. BWR is one type of the most extensively built reactors in the world, with abundant operation experience and mature technologies. Breeder BWR is considered to be capable of being incorporated into the current BWR plants with a handful of modifications, thus obtaining optimal economy.
{"title":"A Core Design of Innovative Breeder BWR","authors":"Guo Rui, A. Yamaji, Yun Cai, Xin-sheng Peng","doi":"10.1115/ICONE26-82079","DOIUrl":"https://doi.org/10.1115/ICONE26-82079","url":null,"abstract":"High breeding with light water cooling has been studied for decades, though is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. To harden the neutron spectrum and thereby to enhance the fuel utilization or even to achieve breeding with light water cooling, the tight-lattice assembly was proposed and applied to High Conversion LWRs. Nonetheless, none of them achieved high breeding. Until recently, the tightly packed fuel assembly (TPFA) is designed for the purpose of high breeding. The ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to be less than 0.1. Super Fast Breeding Reactor (Super FBR) adopts TPFA and achieves breeding performance with compound system doubling time (CSDT) of 43 years. In this study, the breeder BWR core also applies TPFA and achieves CSDT of 50 years. BWR is one type of the most extensively built reactors in the world, with abundant operation experience and mature technologies. Breeder BWR is considered to be capable of being incorporated into the current BWR plants with a handful of modifications, thus obtaining optimal economy.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"62 1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87734423","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 (NPP) have multiple levels of defense in depth hierarchy. The NPP accident condition operation strategy belongs to the 3rd level. It is used to supervise the operator to handle the NPP under accident operating condition. NPP accident condition operation strategy is an essential and difficult work in NPP design field, hence only few organizations are able to develop the accident operating strategies independently all over the world. In this paper, a systematic NPP accident operating condition strategy design methodology is raised based on function analysis and task analysis technology. Based on the methodology, a reactivity accident operation strategy is designed and proved to be reasonable through preliminary verification and validation work.
{"title":"Study on NPP Reactivity Accident Operating Strategy Design Based on Function Analysis and Task Analysis Technology","authors":"Yu Aimin, Xu Zhao, Du Yu, S. Qian","doi":"10.1115/ICONE26-81478","DOIUrl":"https://doi.org/10.1115/ICONE26-81478","url":null,"abstract":"Nuclear Power Plants (NPP) have multiple levels of defense in depth hierarchy. The NPP accident condition operation strategy belongs to the 3rd level. It is used to supervise the operator to handle the NPP under accident operating condition. NPP accident condition operation strategy is an essential and difficult work in NPP design field, hence only few organizations are able to develop the accident operating strategies independently all over the world. In this paper, a systematic NPP accident operating condition strategy design methodology is raised based on function analysis and task analysis technology. Based on the methodology, a reactivity accident operation strategy is designed and proved to be reasonable through preliminary verification and validation work.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"5 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90404590","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}
M. Cappelli, V. Lopresto, Riccardo Cecchi, G. Marrocco
The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.
{"title":"Evaluation of Electromagnetic Fields From Wireless Technologies in a Nuclear Plant","authors":"M. Cappelli, V. Lopresto, Riccardo Cecchi, G. Marrocco","doi":"10.1115/ICONE26-82290","DOIUrl":"https://doi.org/10.1115/ICONE26-82290","url":null,"abstract":"The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"12 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78310960","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}
Compared with the land-based nuclear power plant, the operating conditions of offshore nuclear power plant (ONPP) are much more complicated. For example, the barge-mounted platform malfunction, which is as important as the natural events and human events, should be considered in the plant safety analysis,. As a result, a two dimension operating condition coupled with barge and reactor status should be considered in the development of relevant power plant operating procedures. On the other hand, the beyond design basis hazards induced by the combination of unique and unanticipated external events of ONPP may lead to a blind area to both traditional and two dimension procedures mentioned above. Due to the insufficiency of existing operating condition and relevant procedures to tackle with the above events mentioned, an expanded operation strategy, namely the beyond design basis hazards and the extended ultimate response measures, is developed, Injecting sea water into reactor pressure vessel directly after primary system depressurized and venting the containment when necessary, formed the basis of ultimate response measure, which was proposed by Taiwan Power Company after Fukushima Accident. Considering the offshore and barge-mounted features, the ultimate response measure can be extended to include sea water injection into steam generator indirectly through secondary side passive residual heat removal lines and reactor cabin flooding by sea water through Kingston valves, to rebuild a newly, hierarchical one. Finally, the extended ultimate response measures, provided mainly for the plant command staff and operators, are analyzed utilizing thermal-hydraulic integral computer code preliminarily, to prove the effectiveness of the system configuration and operating strategy. It is concluded that injecting sea water into steam generator can remove the decay heat effectively, and the sensitivity study shows that operator intervention is good enough in accident mitigation.
{"title":"Extended Ultimate Response Measures for Offshore Nuclear Power Plant Under Barge-Reactor Coupled Conditions","authors":"Jue Wang, Longze Li, Chen Hu, W. Cong","doi":"10.1115/ICONE26-81159","DOIUrl":"https://doi.org/10.1115/ICONE26-81159","url":null,"abstract":"Compared with the land-based nuclear power plant, the operating conditions of offshore nuclear power plant (ONPP) are much more complicated. For example, the barge-mounted platform malfunction, which is as important as the natural events and human events, should be considered in the plant safety analysis,. As a result, a two dimension operating condition coupled with barge and reactor status should be considered in the development of relevant power plant operating procedures. On the other hand, the beyond design basis hazards induced by the combination of unique and unanticipated external events of ONPP may lead to a blind area to both traditional and two dimension procedures mentioned above. Due to the insufficiency of existing operating condition and relevant procedures to tackle with the above events mentioned, an expanded operation strategy, namely the beyond design basis hazards and the extended ultimate response measures, is developed, Injecting sea water into reactor pressure vessel directly after primary system depressurized and venting the containment when necessary, formed the basis of ultimate response measure, which was proposed by Taiwan Power Company after Fukushima Accident. Considering the offshore and barge-mounted features, the ultimate response measure can be extended to include sea water injection into steam generator indirectly through secondary side passive residual heat removal lines and reactor cabin flooding by sea water through Kingston valves, to rebuild a newly, hierarchical one. Finally, the extended ultimate response measures, provided mainly for the plant command staff and operators, are analyzed utilizing thermal-hydraulic integral computer code preliminarily, to prove the effectiveness of the system configuration and operating strategy. It is concluded that injecting sea water into steam generator can remove the decay heat effectively, and the sensitivity study shows that operator intervention is good enough in accident mitigation.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"21 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75122945","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}
Xing Mian, Meng Zhaocan, Liao Xiaotao, Sun Canhui, Zhang Shuming, Chen Yaodong, Xiao Hu, Sun Peidong, Huijing Jiang
SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components. A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.
{"title":"Preliminary LOCA Analysis of Heating-Reactor of Advanced Low-Pressurized and Passive Safety System (HAPPY)","authors":"Xing Mian, Meng Zhaocan, Liao Xiaotao, Sun Canhui, Zhang Shuming, Chen Yaodong, Xiao Hu, Sun Peidong, Huijing Jiang","doi":"10.1115/ICONE26-81271","DOIUrl":"https://doi.org/10.1115/ICONE26-81271","url":null,"abstract":"SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components.\u0000 A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"56 5","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72465787","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}
Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.
{"title":"The Canadian Nuclear Safety Commission: Readiness Activities to Regulate Small Modular Reactors","authors":"Kevin Lee","doi":"10.1115/ICONE26-82620","DOIUrl":"https://doi.org/10.1115/ICONE26-82620","url":null,"abstract":"Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"67 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79092923","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 200 MWth nuclear heat reactor II (NHR200-II) is a typical integral pressurized water reactor (iPWR) being developed by the institute of nuclear and new energy technology (INET) in Tsinghua university. The NHR200-II, which has inherent safety features such as full-range natural circulation, passive residual heat removal, self-pressurization and control rod hydraulically driving, can be adopted as a clean base-load energy source for a sea-water desalination plant having the process of multi-effect desalination with thermal vapor compression (MED-TVC). Dynamic modelling of the sea-water desalination plant coupled by the NHR200-II and MED-TVC is necessary for the design of its plant control strategy, which is important for the stable and efficient operation. In this paper, a lumped parameter dynamic model of NHR200II-based sea-water desalination plant with the process of MED-TVC is proposed based upon the conservation laws of mass, momentum and energy. The modeling verification in both the steady-state and open-loop dynamic-state are given, which show the suitability of applying this model for control system design. Finally, the closed-loop responses in the case of power-level maneuver from 100% to 50% full power is given.
{"title":"Dynamic Model of a Seawater Desalination Plant Based on the Nuclear Heating Reactor and MED-TVC","authors":"Yifei Pan, Z. Dong","doi":"10.1115/ICONE26-82556","DOIUrl":"https://doi.org/10.1115/ICONE26-82556","url":null,"abstract":"The 200 MWth nuclear heat reactor II (NHR200-II) is a typical integral pressurized water reactor (iPWR) being developed by the institute of nuclear and new energy technology (INET) in Tsinghua university. The NHR200-II, which has inherent safety features such as full-range natural circulation, passive residual heat removal, self-pressurization and control rod hydraulically driving, can be adopted as a clean base-load energy source for a sea-water desalination plant having the process of multi-effect desalination with thermal vapor compression (MED-TVC). Dynamic modelling of the sea-water desalination plant coupled by the NHR200-II and MED-TVC is necessary for the design of its plant control strategy, which is important for the stable and efficient operation. In this paper, a lumped parameter dynamic model of NHR200II-based sea-water desalination plant with the process of MED-TVC is proposed based upon the conservation laws of mass, momentum and energy. The modeling verification in both the steady-state and open-loop dynamic-state are given, which show the suitability of applying this model for control system design. Finally, the closed-loop responses in the case of power-level maneuver from 100% to 50% full power is given.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"6 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81959292","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}
Takashi Sato, Keiji Matsumoto, K. Hosomi, K. Taguchi
iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the only Generation III.7 reactor incorporating Fukushima lessons learned and complying with Western European Nuclear Regulation Association (WENRA) safety objectives. It is about twice safer than any existing Gen III.5 reactors. It has 7-day grace period for SBO and SA without containment venting. It enables no evacuation and no long-term relocation in SA. It, however, is based on the well-established proven ABWR. The NSSS and TI are exactly the same as those of the existing ABWR. The iB1350 only enhanced the ABWR safety by adding an outer well (OW) as additional PCV volume, built-in passive safety systems (BiPSS) for SA, DEC systems and an APC shield dome over the containment. The BiPSS include an isolation condenser (IC), an innovative passive containment cooling system (iPCCS), in-containment filtered venting system (IFVS), and innovative core catcher (iCC). All the BiPSS are embedded and protected in the containment building against APC. No specialized safety features remote from the R/B are necessary, which reduces plant cost. The primary system has only one integrated RPV. There are no SGs, no pressurizer, no core makeup tanks, no accumulators, no hot legs, and no cold legs. The iB1350 consists of only one integrated RPV and passive safety systems inside the containment building. This configuration is simpler than the simplest large PWR and as simple as SMR. While SMR have rather small outputs, the iB1350 has 1350 MWe output. It is simple, large and economic. As for the safety design it has an in-depth hybrid safety system (IDHS). The IDHS consists of 4 division active safety systems for DBA, 1 or 2 division active safety systems for DEC and the built-in passive safety systems (BiPSS) for SA. The IDHS is originally based on the four levels of safety that have provided an explicit fourth defense level against devastating external events even before 3.11. It also can be explained along with WENRA Defense in Depth (DiD). It is said that independence between DiD levels are important. However, there are many exceptions for independence between DiD levels. For example, SCRAM is used in DiD2, DiD3a and DiD3b. Any DiD that allows exceptions of independence of DiD levels is fake. The iB1350 is rather based on the three levels of safety proposed by Clifford Beck (AEC, 1967). There is complete independence between level 2 (core systems) and level 3 (containment systems) without any exceptions of independence. DiD without exceptions of independence is a real DiD. Only passive safety reactors can meet the real DiD.
{"title":"iB1350: Part 1 — A Generation III.7 Reactor iB1350 and Defense in Depth (DiD)","authors":"Takashi Sato, Keiji Matsumoto, K. Hosomi, K. Taguchi","doi":"10.1115/ICONE26-82428","DOIUrl":"https://doi.org/10.1115/ICONE26-82428","url":null,"abstract":"iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the only Generation III.7 reactor incorporating Fukushima lessons learned and complying with Western European Nuclear Regulation Association (WENRA) safety objectives. It is about twice safer than any existing Gen III.5 reactors. It has 7-day grace period for SBO and SA without containment venting. It enables no evacuation and no long-term relocation in SA. It, however, is based on the well-established proven ABWR. The NSSS and TI are exactly the same as those of the existing ABWR. The iB1350 only enhanced the ABWR safety by adding an outer well (OW) as additional PCV volume, built-in passive safety systems (BiPSS) for SA, DEC systems and an APC shield dome over the containment. The BiPSS include an isolation condenser (IC), an innovative passive containment cooling system (iPCCS), in-containment filtered venting system (IFVS), and innovative core catcher (iCC). All the BiPSS are embedded and protected in the containment building against APC. No specialized safety features remote from the R/B are necessary, which reduces plant cost. The primary system has only one integrated RPV. There are no SGs, no pressurizer, no core makeup tanks, no accumulators, no hot legs, and no cold legs. The iB1350 consists of only one integrated RPV and passive safety systems inside the containment building. This configuration is simpler than the simplest large PWR and as simple as SMR. While SMR have rather small outputs, the iB1350 has 1350 MWe output. It is simple, large and economic. As for the safety design it has an in-depth hybrid safety system (IDHS). The IDHS consists of 4 division active safety systems for DBA, 1 or 2 division active safety systems for DEC and the built-in passive safety systems (BiPSS) for SA. The IDHS is originally based on the four levels of safety that have provided an explicit fourth defense level against devastating external events even before 3.11. It also can be explained along with WENRA Defense in Depth (DiD). It is said that independence between DiD levels are important. However, there are many exceptions for independence between DiD levels. For example, SCRAM is used in DiD2, DiD3a and DiD3b. Any DiD that allows exceptions of independence of DiD levels is fake. The iB1350 is rather based on the three levels of safety proposed by Clifford Beck (AEC, 1967). There is complete independence between level 2 (core systems) and level 3 (containment systems) without any exceptions of independence. DiD without exceptions of independence is a real DiD. Only passive safety reactors can meet the real DiD.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"5 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82388686","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}
H. Yi, Zhang Tian-yi, Wang Jun, YU Yu-Chen, D. Xin
SG (steam generator) is one of the most important equipment in fast reactors, the experience in design and operation of fast reactor worldwide show that failures of SG occurred frequently and often caused serious consequences, therefore it’s necessary to conduct reliability analysis on SG in design phase. FMEA (Failure Mode Effect Analysis) is used to identify all potential failure modes and filter out main failure modes. Then, qualitative analysis and quantitative calculation are carried out to evaluate main failure modes. Next, reliability of SG can be obtained by conducting Latin Hypercube Sampling. Analysis results show that the leakage probability of SG in 20 years is 0.130 219, and the most sensitive factor is the quality of weld in the junction of tubes and tube plate, and the SG meet its reliability requirement.
{"title":"Reliability Evaluation for Steam Generator in a Sodium-Cooled Fast Reactor","authors":"H. Yi, Zhang Tian-yi, Wang Jun, YU Yu-Chen, D. Xin","doi":"10.1115/ICONE26-81183","DOIUrl":"https://doi.org/10.1115/ICONE26-81183","url":null,"abstract":"SG (steam generator) is one of the most important equipment in fast reactors, the experience in design and operation of fast reactor worldwide show that failures of SG occurred frequently and often caused serious consequences, therefore it’s necessary to conduct reliability analysis on SG in design phase. FMEA (Failure Mode Effect Analysis) is used to identify all potential failure modes and filter out main failure modes. Then, qualitative analysis and quantitative calculation are carried out to evaluate main failure modes. Next, reliability of SG can be obtained by conducting Latin Hypercube Sampling. Analysis results show that the leakage probability of SG in 20 years is 0.130 219, and the most sensitive factor is the quality of weld in the junction of tubes and tube plate, and the SG meet its reliability requirement.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"19 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90762900","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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