Benito Mignacca, G. Locatelli, Mahmoud Alaassar, D. Invernizzi
The key characteristics of small modular reactors (SMRs), as their name emphasized, are their size and modularity. Since SMRs are a family of novel reactor designs, there is a gap of empirical knowledge about the cost/benefit analysis of modularization. Conversely, in other sectors (e.g. Oil & Gas) the empirical experience on modularization is much greater. This paper provides a structured knowledge transfer from the general literature (i.e. other major infrastructure) and the Oil & Gas sector to the nuclear power plant construction world. Indeed, in the project management literature, a number of references discuss the costs and benefits determined by the transition from the stick-built construction to modularization, and the main benefits presented in the literature are the reduction of the construction cost and the schedule compression. Additional costs might arise from an increased management hurdle and higher transportation expenses. The paper firstly provides a structured literature review of the benefits and costs of modularization divided into qualitative and quantitative references. In the second part, the paper presents the results of series of interviews with Oil & Gas project managers about the value of modularization in this sector.
{"title":"We Never Built Small Modular Reactors (SMRs), but What Do We Know About Modularization in Construction?","authors":"Benito Mignacca, G. Locatelli, Mahmoud Alaassar, D. Invernizzi","doi":"10.1115/ICONE26-81604","DOIUrl":"https://doi.org/10.1115/ICONE26-81604","url":null,"abstract":"The key characteristics of small modular reactors (SMRs), as their name emphasized, are their size and modularity. Since SMRs are a family of novel reactor designs, there is a gap of empirical knowledge about the cost/benefit analysis of modularization. Conversely, in other sectors (e.g. Oil & Gas) the empirical experience on modularization is much greater. This paper provides a structured knowledge transfer from the general literature (i.e. other major infrastructure) and the Oil & Gas sector to the nuclear power plant construction world. Indeed, in the project management literature, a number of references discuss the costs and benefits determined by the transition from the stick-built construction to modularization, and the main benefits presented in the literature are the reduction of the construction cost and the schedule compression. Additional costs might arise from an increased management hurdle and higher transportation expenses. The paper firstly provides a structured literature review of the benefits and costs of modularization divided into qualitative and quantitative references. In the second part, the paper presents the results of series of interviews with Oil & Gas project managers about the value of modularization in this sector.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"13 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87468330","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}
X. Kong, Yuan Fu, Jianyu Zhang, Hui-Ju Lu, Naxiu Wang
A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m3/h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.
{"title":"Upgrade and Shakedown Test of a High Temperature Fluoride Salt Test Loop","authors":"X. Kong, Yuan Fu, Jianyu Zhang, Hui-Ju Lu, Naxiu Wang","doi":"10.1115/ICONE26-81222","DOIUrl":"https://doi.org/10.1115/ICONE26-81222","url":null,"abstract":"A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m3/h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"139 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86263626","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}
Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.
{"title":"Coordinated Control of a Small Pressurized Water Reactor","authors":"Peiwei Sun, Chongwu Wang","doi":"10.1115/ICONE26-81156","DOIUrl":"https://doi.org/10.1115/ICONE26-81156","url":null,"abstract":"Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"114 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85498750","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}
Chen Lei, Jia Zhen, W. Cong, Gong Zili, Liao Yi, Hu Chen
From the view of practical engineering application, a compacter nuclear power plant is expected. The weight and the volume of a nuclear power plant can be reduced by optimal selection of the operational parameters. In this work, a thermal-hydraulic model of the reactor, mathematical models of the reactor vessel, the main pipe, the pressurizer, the steam generator, the turbine and the condenser were established for the Qinshan-I nuclear power plant based on the related technical materials. The responses of the optimal targets to the changes of the design variables were studied by the sensitivity analyses. The non-dominated solution front of the nuclear power plant was obtained by means of the immune memory clone constrained multi-objective optimization algorithm. The study shows that the component mathematical models are reliable for the optimization process, the distribution of the non-dominated solution is decided by the steam generator secondary pressure. The volume and the weight of the system could be at least reduced by 23.0% and 9.5%, respectively.
{"title":"A Multi-Objective Optimization of the Reactor Power Plant","authors":"Chen Lei, Jia Zhen, W. Cong, Gong Zili, Liao Yi, Hu Chen","doi":"10.1115/ICONE26-82239","DOIUrl":"https://doi.org/10.1115/ICONE26-82239","url":null,"abstract":"From the view of practical engineering application, a compacter nuclear power plant is expected. The weight and the volume of a nuclear power plant can be reduced by optimal selection of the operational parameters. In this work, a thermal-hydraulic model of the reactor, mathematical models of the reactor vessel, the main pipe, the pressurizer, the steam generator, the turbine and the condenser were established for the Qinshan-I nuclear power plant based on the related technical materials. The responses of the optimal targets to the changes of the design variables were studied by the sensitivity analyses. The non-dominated solution front of the nuclear power plant was obtained by means of the immune memory clone constrained multi-objective optimization algorithm. The study shows that the component mathematical models are reliable for the optimization process, the distribution of the non-dominated solution is decided by the steam generator secondary pressure. The volume and the weight of the system could be at least reduced by 23.0% and 9.5%, respectively.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"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":"89650334","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}
Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MWth is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.
{"title":"Design and Feasibility Analysis of the Electricity Generation System Based on Residual Heat","authors":"Z. Dong, Miao Liu, Yifei Pan","doi":"10.1115/ICONE26-82558","DOIUrl":"https://doi.org/10.1115/ICONE26-82558","url":null,"abstract":"Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MWth is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.","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":"85862487","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}
K. Ichikawa, H. Kanda, N. Yoshioka, K. Ara, Jun-ichi Saito, K. Nagai
Studies on the suppression of the reactivity of sodium itself have been performed on the basis of the concept of suspended nanoparticles in liquid sodium (sodium nanofluid). According to the theoretical and experimental results of studies for sodium nanofluid, velocity and heat of sodium nanofluid reaction with water (sodium nanofluid/water reaction) are lower than those of the pure sodium/water reaction. The analytical model for the peak temperature of a sodium nanofluid/water reaction jet has been developed by the authors in consideration of these suppression effects. In this paper, the prediction method for mitigation effects on damage of adjacent tubes in steam generator tube rupture (SGTR) accidents is developed by applying this analytical model for the peak temperature of the reaction jet. On the assumption that the sodium nanofluid is used for the secondary coolant of sodium-cooled fast reactor (SFR), mitigation effects under the design basis accident (DBA) condition and the design extension condition (DEC) of SGTR are estimated by using this method. The results indicate a clear possibility to reduce the number of damaged tubes and to suppress the pressure generated in SGTR accidents by using sodium nanofluid as the secondary coolant.
{"title":"Estimation of Mitigation Effects of Sodium Nanofluid for SGTR Accidents in SFR","authors":"K. Ichikawa, H. Kanda, N. Yoshioka, K. Ara, Jun-ichi Saito, K. Nagai","doi":"10.1115/ICONE26-81309","DOIUrl":"https://doi.org/10.1115/ICONE26-81309","url":null,"abstract":"Studies on the suppression of the reactivity of sodium itself have been performed on the basis of the concept of suspended nanoparticles in liquid sodium (sodium nanofluid). According to the theoretical and experimental results of studies for sodium nanofluid, velocity and heat of sodium nanofluid reaction with water (sodium nanofluid/water reaction) are lower than those of the pure sodium/water reaction. The analytical model for the peak temperature of a sodium nanofluid/water reaction jet has been developed by the authors in consideration of these suppression effects. In this paper, the prediction method for mitigation effects on damage of adjacent tubes in steam generator tube rupture (SGTR) accidents is developed by applying this analytical model for the peak temperature of the reaction jet. On the assumption that the sodium nanofluid is used for the secondary coolant of sodium-cooled fast reactor (SFR), mitigation effects under the design basis accident (DBA) condition and the design extension condition (DEC) of SGTR are estimated by using this method. The results indicate a clear possibility to reduce the number of damaged tubes and to suppress the pressure generated in SGTR accidents by using sodium nanofluid as the secondary coolant.","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":"73425112","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}
Light-water cooled Small Modular Reactors (SMRs) are a potential game-changing technology for energy supply. The potential benefits of SMRs are however conditional on solving the key standardisation and construction issues that have troubled large reactor (LR) projects, which have in turn led to high build costs and long project durations. Initiatives to determine the build schedule of SMRs are hindered by a lack of SMR construction experience and related data. The methodology used in this paper, to deal with the lack of SMR-specific data, draws conclusions about SMRs based on data from actual large pressurised water reactor (PWR) construction experience. It is expected that SMR build schedules can be greatly reduced because of the smaller physical size of structures, fewer components, and other size-related features. However, the construction work space will be more constrained, which could negatively impact build durations. As a result, simple geometric scaling and reduction arguments cannot necessarily be applied to SMR schedules. This paper defines the key areas in which SMR construction differs from LRs, such as smaller geometries as well as modularised and standardised build processes, and describes how these differences might be expected to impact build duration quantitatively. The model developed in this paper presents an approach to determining SMR build schedule durations for a range of reactor sizes. It starts with an LR build schedule based on real data from the UK’s only PWR, Sizewell B. The available data are used to establish a reference point for a non-modular, stick-built SMR schedule. This scheduling approach assumes that, for each major element, part of the time spent on fabrication and installation tasks will vary with reactor size while the remaining fraction will remain constant regardless of reactor size (e.g. due to quality assurance and commissioning tasks). The accuracy of the model generated here is assessed against available construction data and models from a range of actual reactor build projects. The objective of this work is to consider how modularisation can reduce build schedule of SMRs of varying size, by employing modular design and construction principles to both remove tasks that are of long duration from the critical path and to improve construction productivity. Mechanisms by which modularisation reduces build schedule are investigated. Build reduction scenarios are presented based on analysis and subsequent modularisation of the SMR critical path and are compared with other related analyses.
{"title":"A Methodology to Determine SMR Build Schedule and the Impact of Modularisation","authors":"C. Lloyd, A. Roulstone","doi":"10.1115/ICONE26-81550","DOIUrl":"https://doi.org/10.1115/ICONE26-81550","url":null,"abstract":"Light-water cooled Small Modular Reactors (SMRs) are a potential game-changing technology for energy supply. The potential benefits of SMRs are however conditional on solving the key standardisation and construction issues that have troubled large reactor (LR) projects, which have in turn led to high build costs and long project durations.\u0000 Initiatives to determine the build schedule of SMRs are hindered by a lack of SMR construction experience and related data. The methodology used in this paper, to deal with the lack of SMR-specific data, draws conclusions about SMRs based on data from actual large pressurised water reactor (PWR) construction experience.\u0000 It is expected that SMR build schedules can be greatly reduced because of the smaller physical size of structures, fewer components, and other size-related features. However, the construction work space will be more constrained, which could negatively impact build durations. As a result, simple geometric scaling and reduction arguments cannot necessarily be applied to SMR schedules. This paper defines the key areas in which SMR construction differs from LRs, such as smaller geometries as well as modularised and standardised build processes, and describes how these differences might be expected to impact build duration quantitatively.\u0000 The model developed in this paper presents an approach to determining SMR build schedule durations for a range of reactor sizes. It starts with an LR build schedule based on real data from the UK’s only PWR, Sizewell B. The available data are used to establish a reference point for a non-modular, stick-built SMR schedule. This scheduling approach assumes that, for each major element, part of the time spent on fabrication and installation tasks will vary with reactor size while the remaining fraction will remain constant regardless of reactor size (e.g. due to quality assurance and commissioning tasks). The accuracy of the model generated here is assessed against available construction data and models from a range of actual reactor build projects.\u0000 The objective of this work is to consider how modularisation can reduce build schedule of SMRs of varying size, by employing modular design and construction principles to both remove tasks that are of long duration from the critical path and to improve construction productivity. Mechanisms by which modularisation reduces build schedule are investigated. Build reduction scenarios are presented based on analysis and subsequent modularisation of the SMR critical path and are compared with other related analyses.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"68 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78203496","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}
After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.
{"title":"Characteristic Tests on Transition Core of HTR-10","authors":"Liqiang Wei, Dongmei Ding, Ling Liu, Yucheng Wang, Xiaoming Chen, F. Xie","doi":"10.1115/ICONE26-81797","DOIUrl":"https://doi.org/10.1115/ICONE26-81797","url":null,"abstract":"After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"27 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83307602","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 United Kingdom (UK) Small Modular Reactor (SMR) is being developed by a Rolls-Royce led consortium to provide a market driven, affordable, low carbon energy, generation capability. The UK SMR is a Pressurised Water Reactor (PWR) design based on proven technology with a high level of safety achieved through multiple active and passive systems. This paper presents the approach that has been taken in the early design phases of the pressure vessels for the UK SMR. It considers the key design principles e.g. standardisation, simplification and design for manufacture, inspection and assembly which are being applied to enable the cost and lead-time reductions which are necessary for the UK SMR to be a viable alternative to larger conventional nuclear plants. The Reactor Pressure Vessel (RPV) is used as an example to illustrate some of the key design requirements which need to be addressed. Nuclear components are required to be designed and constructed to standards which are commensurate with the significance of the safety functions which they perform. This paper covers the practice established in the UK of designing to Incredibility of Failure for those components with catastrophic failure modes such as the RPV. It describes the additional features including more stringent materials specification and testing, additional defect tolerance studies and the qualification of manufacturing inspections which need to be addressed in the design to satisfy the high reliability claim.
{"title":"Approach to UK SMR Component Design","authors":"C. Bell","doi":"10.1115/ICONE26-81188","DOIUrl":"https://doi.org/10.1115/ICONE26-81188","url":null,"abstract":"The United Kingdom (UK) Small Modular Reactor (SMR) is being developed by a Rolls-Royce led consortium to provide a market driven, affordable, low carbon energy, generation capability. The UK SMR is a Pressurised Water Reactor (PWR) design based on proven technology with a high level of safety achieved through multiple active and passive systems. This paper presents the approach that has been taken in the early design phases of the pressure vessels for the UK SMR. It considers the key design principles e.g. standardisation, simplification and design for manufacture, inspection and assembly which are being applied to enable the cost and lead-time reductions which are necessary for the UK SMR to be a viable alternative to larger conventional nuclear plants. The Reactor Pressure Vessel (RPV) is used as an example to illustrate some of the key design requirements which need to be addressed. Nuclear components are required to be designed and constructed to standards which are commensurate with the significance of the safety functions which they perform. This paper covers the practice established in the UK of designing to Incredibility of Failure for those components with catastrophic failure modes such as the RPV. It describes the additional features including more stringent materials specification and testing, additional defect tolerance studies and the qualification of manufacturing inspections which need to be addressed in the design to satisfy the high reliability claim.","PeriodicalId":65607,"journal":{"name":"International Journal of Plant Engineering and Management","volume":"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":"76529999","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. Tamura, M. Miki, N. Kono, H. Okazawa, S. Okido, C. Zhong, E. Fabre, A. Croxford, P. Wilcox
In power plants, there are structures made up of thin plates, such as air-conditioning ducts or thin-walled pipes, where corrosion can occur. In this study, we provide a solution to reduce inspection time of the thin plate corrosion measurement and enable monitoring, using a non-contact ultrasonic sensor. The sensor can measure the reduction in thickness of thin plates due to general corrosion without the need to remove or reinstall insulating material that is on the outside of the plate. The proposed sensor is based on the non-contact ultrasonic measurement technique which was originally proposed by Greve et al, further developed and patented by Zhong et al. at the University of Bristol, and commercialized by Inductosense Ltd. In order to ultrasonically measure the thin plate thickness, we use a method based on the group velocity of the guided waves. The proposed method was tested theoretically with numerical simulations and experimentally against our target conditions. The results of the numerical simulations and experiments confirm that the proposed method can be applied to thickness measurements of thin-plates in our target condition. Based on the feasibility test results, we developed a prototype sensor and measurement software. From the results of the performance evaluation tests, we have confirmed that the prototype sensor has sufficient capability to measure the thickness of the thin plates without the removal of the insulator. Even if the offset between the plate and the inspection probe is 100 mm, the prototype sensor still works well.
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