The purpose of tokamak plasma diagnostics is to provide the necessary parameters for device protection, operation, and maintenance. It can also supply parameters for fusion physics research. As one of the main ways to diagnose nuclear fusion plasma, neutron diagnosis focuses on the detection of neutrons, produced by the D-D and D-T fusion reactions, to obtain the physical information of internal plasma. Neutron measurements are widely performed on tokamak to provide the essential information on the neutron yield rate of the plasma that is related to fusion power. Since neutron has no electric charge, neutron can’t be ionized directly by the interaction of electrons in the detection material. The interactions between neutron and nuclei, such as nuclear reaction and nuclear recoil, are used to detect neutrons. According to the front sensitive materials, neutron detectors can be divided into gas detectors, scintillator detectors, semiconductor detectors, ionization chambers and so on. Since the magnetic field surrounding Tokamak can have a magnificent influence on the performance of photo-electronic multiplier tubes (PMTs), it is necessary to employ magnetic shielding in designing detectors, thus guaranteeing the proper operation of detectors within a strong magnetic field. Although the PMTs are equipped with magnetic shielding materials by manufacturers, they can only resist the influence of geomagnetic field. Besides the magnetic shielding and neutron/gamma shielding, neutron detectors should be calibrated before used on the tokamak. Nine similar detectors were assembled and calibrated in this paper. The basic idea of processing calibration data is that we should adjust the resolution and the light response function in order to make experiment spectrum and simulation spectrum fit on the recoil proton edge. A special explication is given to the data processing of neutron calibration, followed by an analysis of its resulting light response function and by comparison with PTB’s results.
{"title":"Energy Calibration of Scintillator Detectors in Different Neutron Diagnostic System on Tokamak","authors":"Z. Cui","doi":"10.1115/ICONE26-81190","DOIUrl":"https://doi.org/10.1115/ICONE26-81190","url":null,"abstract":"The purpose of tokamak plasma diagnostics is to provide the necessary parameters for device protection, operation, and maintenance. It can also supply parameters for fusion physics research. As one of the main ways to diagnose nuclear fusion plasma, neutron diagnosis focuses on the detection of neutrons, produced by the D-D and D-T fusion reactions, to obtain the physical information of internal plasma. Neutron measurements are widely performed on tokamak to provide the essential information on the neutron yield rate of the plasma that is related to fusion power. Since neutron has no electric charge, neutron can’t be ionized directly by the interaction of electrons in the detection material. The interactions between neutron and nuclei, such as nuclear reaction and nuclear recoil, are used to detect neutrons. According to the front sensitive materials, neutron detectors can be divided into gas detectors, scintillator detectors, semiconductor detectors, ionization chambers and so on. Since the magnetic field surrounding Tokamak can have a magnificent influence on the performance of photo-electronic multiplier tubes (PMTs), it is necessary to employ magnetic shielding in designing detectors, thus guaranteeing the proper operation of detectors within a strong magnetic field. Although the PMTs are equipped with magnetic shielding materials by manufacturers, they can only resist the influence of geomagnetic field. Besides the magnetic shielding and neutron/gamma shielding, neutron detectors should be calibrated before used on the tokamak. Nine similar detectors were assembled and calibrated in this paper. The basic idea of processing calibration data is that we should adjust the resolution and the light response function in order to make experiment spectrum and simulation spectrum fit on the recoil proton edge. A special explication is given to the data processing of neutron calibration, followed by an analysis of its resulting light response function and by comparison with PTB’s results.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"11 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133170185","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 framework – NuTERA (Nuclear Techno-Economic and Risk Assessment) has been developed to set out the requirements for evaluating Generation IV (Gen IV) Nuclear Power Plants (NPPs) at the design conceptual stage. The purpose of the framework is to provide guidelines for future tools that are required to support the decision-making process on the choice of Gen IV concepts and cycle configurations. In this paper, the underpinning of the framework has been demonstrated to enable the creation of an analyses tool, which evaluates the design of an NPP that utilises helium closed Brayton gas turbine cycles. The tool at the broad spectrum focuses on the component and cycle design, Design Point (DP) and Off-Design Point (ODP) performance, part power and load following operations. Specifically, the design model has been created to provide functionalities that look at the in-depth sensitivities of the design factors and operation that affect the efficiency of an NPP such as temperature and pressure ratios, inlet cycle temperatures, component efficiencies, pressure losses. The ODP performance capabilities include newly derived component maps for the reactor, intercooler and recuperator for long term Off-Design (OD) operation. With regard to short term OD, which is typically driven by changes in ambient conditions, the ability to analyse the cycle load following capabilities are possible. An economic model has also been created, which calculates the component costs and the baseline economic evaluation. An incorporated risk model quantifies the performance, operational, financial and design impact risks. However, the tool is able to optimise the NPP cycle configuration based on the best economics using the Levelised Unit Electricity Cost (LUEC) as a measure. The tool has been used to demonstrate a typical decision-making process on 2 Gen IV helium closed gas turbine cycles, which apply to the Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs). The cycles are the Simple Cycle Recuperator (SCR) and Intercooled Cycle Recuperator (ICR). The tool was able to derive the most efficient cycle configurations for the ICR (53% cycle efficiency) and SCR (50% cycle efficiency). Based on these efficiency figures, the baseline LUEC ($/MWh) for the year 2020 is $62.13 for the ICR and $61.84 for the SCR. However, the inclusion of the cost of contingencies due to risks and the subsequent economic optimisation resulted in a cost of $69.70 and $69.80 for the ICR and SCR respectively.
{"title":"A Framework and Model for Assessing the Design Point Performance, Off-Design Point Performance, Control, Economics and Risks of Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants","authors":"A. Gad-Briggs, P. Pilidis, T. Nikolaidis","doi":"10.1115/ICONE26-81686","DOIUrl":"https://doi.org/10.1115/ICONE26-81686","url":null,"abstract":"A framework – NuTERA (Nuclear Techno-Economic and Risk Assessment) has been developed to set out the requirements for evaluating Generation IV (Gen IV) Nuclear Power Plants (NPPs) at the design conceptual stage. The purpose of the framework is to provide guidelines for future tools that are required to support the decision-making process on the choice of Gen IV concepts and cycle configurations. In this paper, the underpinning of the framework has been demonstrated to enable the creation of an analyses tool, which evaluates the design of an NPP that utilises helium closed Brayton gas turbine cycles. The tool at the broad spectrum focuses on the component and cycle design, Design Point (DP) and Off-Design Point (ODP) performance, part power and load following operations. Specifically, the design model has been created to provide functionalities that look at the in-depth sensitivities of the design factors and operation that affect the efficiency of an NPP such as temperature and pressure ratios, inlet cycle temperatures, component efficiencies, pressure losses. The ODP performance capabilities include newly derived component maps for the reactor, intercooler and recuperator for long term Off-Design (OD) operation. With regard to short term OD, which is typically driven by changes in ambient conditions, the ability to analyse the cycle load following capabilities are possible. An economic model has also been created, which calculates the component costs and the baseline economic evaluation. An incorporated risk model quantifies the performance, operational, financial and design impact risks. However, the tool is able to optimise the NPP cycle configuration based on the best economics using the Levelised Unit Electricity Cost (LUEC) as a measure. The tool has been used to demonstrate a typical decision-making process on 2 Gen IV helium closed gas turbine cycles, which apply to the Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs). The cycles are the Simple Cycle Recuperator (SCR) and Intercooled Cycle Recuperator (ICR). The tool was able to derive the most efficient cycle configurations for the ICR (53% cycle efficiency) and SCR (50% cycle efficiency). Based on these efficiency figures, the baseline LUEC ($/MWh) for the year 2020 is $62.13 for the ICR and $61.84 for the SCR. However, the inclusion of the cost of contingencies due to risks and the subsequent economic optimisation resulted in a cost of $69.70 and $69.80 for the ICR and SCR respectively.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"33 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116575359","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}
Qian Sun, Tianji Peng, Zhiwei Zhou, Zhibin Chen, Shisheng Wang
The use of liquid metal alloy as a coolant in nuclear systems receives more and more attention in recent years, such as Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM) for ITER, Lead-bismuth spallation target for accelerator driven sub-critical system (ADS) and Lead-alloy-cooled fast reactor (LFR). A system safety analysis code is an important tool for the liquid metal nuclear system safety analysis. In order to analyze some of the basis accidents, there is the need to simulate the mixing of liquid metal and noncondensable gas. While the current system safety code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems is incapable of modeling the mixture of liquid metal fluids and noncondensable gas. This paper first briefly introduce the two-fluid model in RELAP5/MOD4.0 and the reason for its incapability of modeling liquid metal in contact with a non-condensable gas. Then, a solution to solve the problem and the modification of the RELAP5/MOD4.0 code is proposed. Last, several typical problems in DFLL-TBM system were simulated and the results demonstrate the feasibility and validity of the modified RELAP5/MOD4.0 in modeling the mixing of liquid metal and non-condensable gas. The modified code provides a powerful tool for liquid metal nuclear system designs and safety analysis.
{"title":"Modification of RELAP/SCDAPSIM/MOD4.0 for Liquid Metal in Contact With Noncondensable Gas","authors":"Qian Sun, Tianji Peng, Zhiwei Zhou, Zhibin Chen, Shisheng Wang","doi":"10.1115/ICONE26-82030","DOIUrl":"https://doi.org/10.1115/ICONE26-82030","url":null,"abstract":"The use of liquid metal alloy as a coolant in nuclear systems receives more and more attention in recent years, such as Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM) for ITER, Lead-bismuth spallation target for accelerator driven sub-critical system (ADS) and Lead-alloy-cooled fast reactor (LFR). A system safety analysis code is an important tool for the liquid metal nuclear system safety analysis. In order to analyze some of the basis accidents, there is the need to simulate the mixing of liquid metal and noncondensable gas. While the current system safety code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems is incapable of modeling the mixture of liquid metal fluids and noncondensable gas. This paper first briefly introduce the two-fluid model in RELAP5/MOD4.0 and the reason for its incapability of modeling liquid metal in contact with a non-condensable gas. Then, a solution to solve the problem and the modification of the RELAP5/MOD4.0 code is proposed. Last, several typical problems in DFLL-TBM system were simulated and the results demonstrate the feasibility and validity of the modified RELAP5/MOD4.0 in modeling the mixing of liquid metal and non-condensable gas. The modified code provides a powerful tool for liquid metal nuclear system designs and safety analysis.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"169 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124724693","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In 2016, IAEA revised and issued SSR-2/1 Safety of Nuclear Power Plants: Design and issued SSR-2/1 (Rev.1). Major revision in SSR-2/1 (Rev.1) is analyzed (especially new requirements after Fukushima nuclear accident). Problems in SSR-2/1 (Rev.1) are also discussed, which could be used as a reference in HAF102 revision and SSR-2/1 (Rev.1) application.
{"title":"Technical Insights of SSR-2/1 Safety of Nuclear Power Plants: Design (Rev.1)","authors":"Zheng Hua, Wei Shuhong","doi":"10.1115/ICONE26-81983","DOIUrl":"https://doi.org/10.1115/ICONE26-81983","url":null,"abstract":"In 2016, IAEA revised and issued SSR-2/1 Safety of Nuclear Power Plants: Design and issued SSR-2/1 (Rev.1). Major revision in SSR-2/1 (Rev.1) is analyzed (especially new requirements after Fukushima nuclear accident). Problems in SSR-2/1 (Rev.1) are also discussed, which could be used as a reference in HAF102 revision and SSR-2/1 (Rev.1) application.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"39 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133661410","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}
G. Giambartolomei, R. L. Frano, Dahmane Mazed, D. Serra, D. Aquaro
The condensation of steam in water may cause pressure oscillations, vibrations, and damages to piping and, in some extreme conditions to the internals of suppression tank. Significant R&D activity, especially focusing on the emergency cooling system in BWRs, has been conducted in the past decades to clarify the mechanism of the condensation oscillation evolving/related to the Direct Contact Condensation (DCC). The present paper deals with the pressure oscillations due to the condensation of steam jet in water, at sub-atmospheric condition; phenomena that have not been fully investigated yet. Vibrations, arisen as flow patterns transformed from stable to unstable, are investigated with particular attention to the dynamic behaviour of the whole suppression system, behavior of the interface, flow patterns etc. To the purpose, a numerical study is performed by means of suitable FEM in order to identify the dominant frequency associated to the steam jet condensation oscillation in water flow, and determine the resulting values of pressure and acceleration. The obtained results allowed to formulate a correlation between the dominant frequency and the condensation driving potential and steam mass flux.
{"title":"Analysis of Vibrations due to the Steam Condensation at Sub-Atmospheric Condition","authors":"G. Giambartolomei, R. L. Frano, Dahmane Mazed, D. Serra, D. Aquaro","doi":"10.1115/ICONE26-82378","DOIUrl":"https://doi.org/10.1115/ICONE26-82378","url":null,"abstract":"The condensation of steam in water may cause pressure oscillations, vibrations, and damages to piping and, in some extreme conditions to the internals of suppression tank. Significant R&D activity, especially focusing on the emergency cooling system in BWRs, has been conducted in the past decades to clarify the mechanism of the condensation oscillation evolving/related to the Direct Contact Condensation (DCC).\u0000 The present paper deals with the pressure oscillations due to the condensation of steam jet in water, at sub-atmospheric condition; phenomena that have not been fully investigated yet. Vibrations, arisen as flow patterns transformed from stable to unstable, are investigated with particular attention to the dynamic behaviour of the whole suppression system, behavior of the interface, flow patterns etc.\u0000 To the purpose, a numerical study is performed by means of suitable FEM in order to identify the dominant frequency associated to the steam jet condensation oscillation in water flow, and determine the resulting values of pressure and acceleration. The obtained results allowed to formulate a correlation between the dominant frequency and the condensation driving potential and steam mass flux.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"39 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133126287","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}
Koji Yamada, Kyoichi Chuda, A. Masu, Shoichi Goto, Isamu Nakazuka, Yohei Komiyama, Tatsumi Horiuchi, Iwata Tomokazu, Tsumura Yasuhiro, Shizuo Noda
“Technical Guidelines for Watertight Facilities (JEAG4630)” was established in 2016 to provide common guidelines for nuclear power utilities. This technical guidance describes the basic processes of each stage, from design to in-service inspections in Japanese. The guidance considers the features of the equipment, the functions of which are difficult to inspect after installation. This guidance is formulated according to the requirements described in Chapter 4 of the “Technical Code for Tsunami Design of Nuclear Power Plants (JEAC4629-2014)”: “Design of Tsunami Protection and Flood Protection Facilities”.
{"title":"Establishment of “Technical Guidelines for Watertight Facilities (JEAG4630-2016)”","authors":"Koji Yamada, Kyoichi Chuda, A. Masu, Shoichi Goto, Isamu Nakazuka, Yohei Komiyama, Tatsumi Horiuchi, Iwata Tomokazu, Tsumura Yasuhiro, Shizuo Noda","doi":"10.1115/ICONE26-81208","DOIUrl":"https://doi.org/10.1115/ICONE26-81208","url":null,"abstract":"“Technical Guidelines for Watertight Facilities (JEAG4630)” was established in 2016 to provide common guidelines for nuclear power utilities. This technical guidance describes the basic processes of each stage, from design to in-service inspections in Japanese. The guidance considers the features of the equipment, the functions of which are difficult to inspect after installation. This guidance is formulated according to the requirements described in Chapter 4 of the “Technical Code for Tsunami Design of Nuclear Power Plants (JEAC4629-2014)”: “Design of Tsunami Protection and Flood Protection Facilities”.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"104 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124645709","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}
Xiao-Yun Wang, Siyang Huang, W. Tian, Lie Chen, S. Qiu, G. Su
In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.
{"title":"Effect of Rolling Motion on Flow Instability of Parallel Rectangular Channels of Natural Circulation","authors":"Xiao-Yun Wang, Siyang Huang, W. Tian, Lie Chen, S. Qiu, G. Su","doi":"10.1115/ICONE26-81849","DOIUrl":"https://doi.org/10.1115/ICONE26-81849","url":null,"abstract":"In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"63 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123794979","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}
Experience feedback refers to the timely information collection, transformation, analysis, processing and summary, when some good experience or occasional problems appeared during the manufacturing process. In the manufacturing process of nuclear fuel, CNNC JianZhong Nuclear Fuel Assembly Co., Ltd (CJNF) established a comprehensive experience feedback system, and consolidated experience feedback processing flow. Using classification and gradation to collect, filter, organize and use information. CJNF feeds back some quality problems in fuel assembly manufacture, through the analysis of causes and the implementation of measures to avoid the occurrence of similar problems. Meantime, feeding back and sharing good practical experience in manufacturing and management process, it is benefit to pass on experience and learn from each other. The experience feedback of CJNF is the solid foundation of quality management system’s operation and improvement.
{"title":"The Quality Experience Feedback in Nuclear Fuel Manufacture","authors":"Tong MengYao, Li FangGang","doi":"10.1115/ICONE26-81374","DOIUrl":"https://doi.org/10.1115/ICONE26-81374","url":null,"abstract":"Experience feedback refers to the timely information collection, transformation, analysis, processing and summary, when some good experience or occasional problems appeared during the manufacturing process. In the manufacturing process of nuclear fuel, CNNC JianZhong Nuclear Fuel Assembly Co., Ltd (CJNF) established a comprehensive experience feedback system, and consolidated experience feedback processing flow. Using classification and gradation to collect, filter, organize and use information. CJNF feeds back some quality problems in fuel assembly manufacture, through the analysis of causes and the implementation of measures to avoid the occurrence of similar problems. Meantime, feeding back and sharing good practical experience in manufacturing and management process, it is benefit to pass on experience and learn from each other. The experience feedback of CJNF is the solid foundation of quality management system’s operation and improvement.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133299183","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}
Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.
{"title":"The Characteristics Study of Helium-Xenon Mixture in Closed Brayton Cycle for Space Nuclear Reactor Power","authors":"Xie Yang, Lei Shi","doi":"10.1115/ICONE26-82220","DOIUrl":"https://doi.org/10.1115/ICONE26-82220","url":null,"abstract":"Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"52 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131588503","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}
During the past 30 years the main rules to design pressure vessels were based on elastic analyses. Many conservatisms associated to these different elastic approaches are discussed in this paper, like: stress criteria linearization for 3-D components, stress classification in nozzle areas, plastic shake down analysis, fatigue analysis, Ke evaluation, and pipe stress criteria for elastic follow-up due to thermal expansion or seismic loads... This paper will improve existing codified rules in nuclear and non-nuclear Codes that are proposed as alternatives to elastic evaluation for different failure modes and degradation mechanisms: plastic collapse, plastic instability, tri-axial local failure, rupture of cracked component, fatigue and Ke, plastic shakedown. These methods are based on limit loads, monotonic or cyclic elastic-plastic analyses. Concerned components are mainly vessels and piping systems. No existing Code is sufficiently detailed to be easily applied; the needs are stress analysis methods through finite elements, material properties including material constitutive equations and criteria associated to each methods and each failure modes. A first set of recommendation to perform these inelastic analysis will be presented to improve existing codes on an international harmonized way, associated to all material properties and criteria needed to apply these modern methods. An international draft Code Case is in preparation.
{"title":"Nonlinear Analysis in Pressure Vessel Design Codes: Recommendations for Codified Rules Improvements","authors":"C. Faidy","doi":"10.1115/ICONE26-81095","DOIUrl":"https://doi.org/10.1115/ICONE26-81095","url":null,"abstract":"During the past 30 years the main rules to design pressure vessels were based on elastic analyses. Many conservatisms associated to these different elastic approaches are discussed in this paper, like: stress criteria linearization for 3-D components, stress classification in nozzle areas, plastic shake down analysis, fatigue analysis, Ke evaluation, and pipe stress criteria for elastic follow-up due to thermal expansion or seismic loads...\u0000 This paper will improve existing codified rules in nuclear and non-nuclear Codes that are proposed as alternatives to elastic evaluation for different failure modes and degradation mechanisms: plastic collapse, plastic instability, tri-axial local failure, rupture of cracked component, fatigue and Ke, plastic shakedown. These methods are based on limit loads, monotonic or cyclic elastic-plastic analyses.\u0000 Concerned components are mainly vessels and piping systems.\u0000 No existing Code is sufficiently detailed to be easily applied; the needs are stress analysis methods through finite elements, material properties including material constitutive equations and criteria associated to each methods and each failure modes.\u0000 A first set of recommendation to perform these inelastic analysis will be presented to improve existing codes on an international harmonized way, associated to all material properties and criteria needed to apply these modern methods. An international draft Code Case is in preparation.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129200944","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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