The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MWth nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.
{"title":"Dynamic Modeling of the NSSS Based on NHR200-II Nuclear Heating Reactor","authors":"Z. Dong, Yifei Pan, Miao Liu, Xiaojin Huang","doi":"10.1115/ICONE26-82579","DOIUrl":"https://doi.org/10.1115/ICONE26-82579","url":null,"abstract":"The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MWth nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.","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":"88452527","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 new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators. Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements. To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.
{"title":"Control Strategy Investigation for a Multi-Purpose Modular Small Pressurized Water Reactor With Once-Through Steam Generators","authors":"Qian Ma, Peiwei Sun","doi":"10.1115/ICONE26-81318","DOIUrl":"https://doi.org/10.1115/ICONE26-81318","url":null,"abstract":"A new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators.\u0000 Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements.\u0000 To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.","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":"83480693","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}
Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO2 cycles can also be designed for operation under direct dry air cooling. This paper presents the results of the coupled control analysis of a sCO2 cycle for a 100 MWe sodium-cooled fast reactor. The plant control mechanisms were investigated and optimized for load following operation.
{"title":"Dynamic Control Analysis of the AFR-100 SMR SFR With a Supercritical CO2 Cycle and Dry Air Cooling: Part I — Plant Control Optimization","authors":"A. Moisseytsev, J. Sienicki","doi":"10.1115/ICONE26-82292","DOIUrl":"https://doi.org/10.1115/ICONE26-82292","url":null,"abstract":"Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO2 cycles can also be designed for operation under direct dry air cooling. This paper presents the results of the coupled control analysis of a sCO2 cycle for a 100 MWe sodium-cooled fast reactor. The plant control mechanisms were investigated and optimized for load following operation.","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":"88240728","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}
There was a common use of instrument and control (I&C) system based on analog technology in design and construction of nuclear power plant built more than ten years ago. With the development and update of automation technology, digital control system has almost completely replaced the older generation technology in many areas of industry. For the nuclear power plant still using the analog I&C system, it caused the reduction of the production lines of related components and the lack of specified technical engineer. Along with the aging and obsolescence of analog technology equipment, the unsustainability of spares prompted the owners of nuclear power plant to implement modernization using digital technologies. Different from the design of digital control system in new nuclear power plant, the modernization project design is limited by the setting of the original system. Therefore, most owners adopt the function alternative strategy to implement the upgrading project. This strategy can effectively solve the problem of spares shortage, but it is difficult to fully elaborate the advantages of digital control system. Based on the technical characteristics of digital control system and the design experience derived from new nuclear power project construction, this paper puts forward the optimized design measures, under the limitation of the old power plant, to enhance the safety and economy of the nuclear power plant after final digital upgrading.
{"title":"Design Optimization of Modernization of I&C System Using Digital Technology in NPPs","authors":"Long-qiang Zhang, Jiahong Yan, Weining Zhao, Weijun Huang","doi":"10.1115/ICONE26-82498","DOIUrl":"https://doi.org/10.1115/ICONE26-82498","url":null,"abstract":"There was a common use of instrument and control (I&C) system based on analog technology in design and construction of nuclear power plant built more than ten years ago. With the development and update of automation technology, digital control system has almost completely replaced the older generation technology in many areas of industry. For the nuclear power plant still using the analog I&C system, it caused the reduction of the production lines of related components and the lack of specified technical engineer. Along with the aging and obsolescence of analog technology equipment, the unsustainability of spares prompted the owners of nuclear power plant to implement modernization using digital technologies. Different from the design of digital control system in new nuclear power plant, the modernization project design is limited by the setting of the original system. Therefore, most owners adopt the function alternative strategy to implement the upgrading project. This strategy can effectively solve the problem of spares shortage, but it is difficult to fully elaborate the advantages of digital control system. Based on the technical characteristics of digital control system and the design experience derived from new nuclear power project construction, this paper puts forward the optimized design measures, under the limitation of the old power plant, to enhance the safety and economy of the nuclear power plant after final digital upgrading.","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":"88990954","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}
Caike Zhang, Jing Qi, C. Liu, Chenglong Xie, Peibang Liu, Ming Qu
At present, DCS is widely used as the control system for nuclear power plants both at home and abroad, which prompting many companies to research the technology of DCS debugging. In this paper, taking a certain nuclear power plant within China for reference, the virtual DCS debugging and research platform which based on the full-scope nuclear power plant simulation model is developed. It was developed by first establishing a simulation model on the RINSIM Simulation Platform and ordering a customized set of virtual DCS system, then developing a communication program between the simulation model and the virtual DCS system. Users can observe the actual effects and results if following the pre-designed testing procedures after the configuration of control logics, HMI (Human Machine Interface) and I/O communication interfaces. The virtual DCS platform is aimed at assisting with technology research of DCS project for similar nuclear power plants and also can provide professional training for associated personnel of nuclear power plant.
{"title":"Design and Development of Virtual DCS Debugging and Research Platform Based on NPP Simulation Model","authors":"Caike Zhang, Jing Qi, C. Liu, Chenglong Xie, Peibang Liu, Ming Qu","doi":"10.1115/ICONE26-81122","DOIUrl":"https://doi.org/10.1115/ICONE26-81122","url":null,"abstract":"At present, DCS is widely used as the control system for nuclear power plants both at home and abroad, which prompting many companies to research the technology of DCS debugging. In this paper, taking a certain nuclear power plant within China for reference, the virtual DCS debugging and research platform which based on the full-scope nuclear power plant simulation model is developed. It was developed by first establishing a simulation model on the RINSIM Simulation Platform and ordering a customized set of virtual DCS system, then developing a communication program between the simulation model and the virtual DCS system. Users can observe the actual effects and results if following the pre-designed testing procedures after the configuration of control logics, HMI (Human Machine Interface) and I/O communication interfaces. The virtual DCS platform is aimed at assisting with technology research of DCS project for similar nuclear power plants and also can provide professional training for associated personnel of nuclear power plant.","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":"87827287","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}
This paper researched the safety functional requirements analysis and the allocation of functions between man and machine for the nuclear power plant. The safety functional requirements are identified from accident handling needs and refined from system configuration consideration. Through the analysis of design conditions, some safety features were extracted to mitigate accidents. Then, components (e.g. pumps, valves, tanks) were determined to implement each of the safety features at the system design stage. At this stage, some implicit safety features, which could not be obtained directly from the accident analysis, were added, according to the specific conditions of system configuration and operation. Finally, after further judgement on possible inconsistency, a complete list of safety functions for the nuclear power plant was formed. As an illustration, this paper provided a list of safety functions related to the safety injection function, and a list of equipment for the safety injection system. Furthermore, these identified safety functions, were appropriately allocated between man and machine, to be performed either by system components automatically, or by operators locally or remotely from the control room, or under the cooperation of operators and system components. Seven factors were considered in the allocation: a) performance requirements; b) the capability or limits of man and machine; c) existing practices; d) operating experience; e) management requirement; f) technical feasibility; g) cost. The allocation of functions for the safety injection system was validated using a simulator.
{"title":"Research on Nuclear Power Plant Safety Functional Requirements Analysis and Function Allocation","authors":"J. Ming, H. Huan, Zhang Xuegang","doi":"10.1115/ICONE26-82230","DOIUrl":"https://doi.org/10.1115/ICONE26-82230","url":null,"abstract":"This paper researched the safety functional requirements analysis and the allocation of functions between man and machine for the nuclear power plant. The safety functional requirements are identified from accident handling needs and refined from system configuration consideration. Through the analysis of design conditions, some safety features were extracted to mitigate accidents. Then, components (e.g. pumps, valves, tanks) were determined to implement each of the safety features at the system design stage. At this stage, some implicit safety features, which could not be obtained directly from the accident analysis, were added, according to the specific conditions of system configuration and operation. Finally, after further judgement on possible inconsistency, a complete list of safety functions for the nuclear power plant was formed. As an illustration, this paper provided a list of safety functions related to the safety injection function, and a list of equipment for the safety injection system. Furthermore, these identified safety functions, were appropriately allocated between man and machine, to be performed either by system components automatically, or by operators locally or remotely from the control room, or under the cooperation of operators and system components. Seven factors were considered in the allocation: a) performance requirements; b) the capability or limits of man and machine; c) existing practices; d) operating experience; e) management requirement; f) technical feasibility; g) cost. The allocation of functions for the safety injection system was validated using a simulator.","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":"77727424","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 nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.
{"title":"Verification of Alarm Displays for the Nuclear Power Plant With Two Modular High-Temperature Gas-Cooled Reactors","authors":"Jiao Qianqian, G. Chao, L. Jianghai, Qu Ronghong","doi":"10.1115/ICONE26-82561","DOIUrl":"https://doi.org/10.1115/ICONE26-82561","url":null,"abstract":"The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.","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":"79335870","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}
Discontinuity appears in simulated training instruments of nuclear radiation reconnaissance which adopted ellipse numerical model. In order to solve the problem, an improved ellipse numerical model was proposed, in which the contaminated area was taken into count as a whole. The level of nuclear radiation at any position in the contaminated area can be calculated by the improved ellipse numerical model. On the basis of the improved ellipse numerical model, the architecture of the simulated instrument for training of nuclear radiation reconnaissance was proposed. The results of experiments showed that the improved ellipse numerical model not only had the main characteristics of the contaminated area but also successfully solved the problem of numerical discontinuity. Through adjusting the parameters of the contaminated area, the improved training instrument can adapt to different scope of nuclear radiation reconnaissance without any regional restriction.
{"title":"Simulated Training Instrument of Nuclear Radiation Reconnaissance Based on an Improved Ellipse Numerical Model","authors":"Shuijun He, Manchun Liang, G. Su, J. T. He","doi":"10.1115/ICONE26-81250","DOIUrl":"https://doi.org/10.1115/ICONE26-81250","url":null,"abstract":"Discontinuity appears in simulated training instruments of nuclear radiation reconnaissance which adopted ellipse numerical model. In order to solve the problem, an improved ellipse numerical model was proposed, in which the contaminated area was taken into count as a whole. The level of nuclear radiation at any position in the contaminated area can be calculated by the improved ellipse numerical model. On the basis of the improved ellipse numerical model, the architecture of the simulated instrument for training of nuclear radiation reconnaissance was proposed. The results of experiments showed that the improved ellipse numerical model not only had the main characteristics of the contaminated area but also successfully solved the problem of numerical discontinuity. Through adjusting the parameters of the contaminated area, the improved training instrument can adapt to different scope of nuclear radiation reconnaissance without any regional restriction.","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":"75020199","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}
Liu Qiaofen, Xiao Sanping, Liu Yu, Liu Xichao, Jiang Xulun
Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.
{"title":"Research on Algorithm of Sump Level Operator Assisted Support Program for PWR Nuclear Power Plant","authors":"Liu Qiaofen, Xiao Sanping, Liu Yu, Liu Xichao, Jiang Xulun","doi":"10.1115/ICONE26-81714","DOIUrl":"https://doi.org/10.1115/ICONE26-81714","url":null,"abstract":"Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.","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":"75045644","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}
S. Haas, D. Chu, Kevin Ellis, M. White, B. Lindley, P. Smith, J. Murgatroyd, A. Grief, Mike Leddy, Mike Yule
Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR. The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material. A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken. This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.
{"title":"Integrated Design of a Reactor Core for the Rolls-Royce Small Modular Reactor Project","authors":"S. Haas, D. Chu, Kevin Ellis, M. White, B. Lindley, P. Smith, J. Murgatroyd, A. Grief, Mike Leddy, Mike Yule","doi":"10.1115/ICONE26-81311","DOIUrl":"https://doi.org/10.1115/ICONE26-81311","url":null,"abstract":"Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR.\u0000 The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material.\u0000 A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken.\u0000 This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.","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":"76133084","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：