A. Zvorykina, D. Khmil, N. Fialko, I. Pioro, Svitlana Stryzheus
In this paper CFD analyses of mixed convection in bare tubes cooled with water at supercritical pressure is presented. The study was carried out using the FLUENT code for upward flow in vertical tubes with a heated length of 4 m and an inside diameter of 10 mm at relatively low water mass flux (G ≈500 kg/m2s) and heat fluxes q (from 239 to 310 kW/m2). Various models of turbulence have been tested. The results of the studies demonstrated a reasonable good agreement between CFD predictions and experimental data on the heat transfer coefficient and internal-wall temperature with use the SST turbulence model. Comparison of the CFD simulation data, which correspond to the presence or absence of the buoyancy forces, was performed. The regularities of the influence of these forces on the damping of turbulent transport, the deformation of the radial profiles of velocity and temperature along the channel length, the reduction of the heat transfer coefficients, etc. were studied. The features of the motion of the pseudo-phase transition within various conditions are presented.
{"title":"CFD Analysis of Supercritical-Water Flow and Heat Transfer in Vertical Bare Tube","authors":"A. Zvorykina, D. Khmil, N. Fialko, I. Pioro, Svitlana Stryzheus","doi":"10.1115/ICONE26-81045","DOIUrl":"https://doi.org/10.1115/ICONE26-81045","url":null,"abstract":"In this paper CFD analyses of mixed convection in bare tubes cooled with water at supercritical pressure is presented. The study was carried out using the FLUENT code for upward flow in vertical tubes with a heated length of 4 m and an inside diameter of 10 mm at relatively low water mass flux (G ≈500 kg/m2s) and heat fluxes q (from 239 to 310 kW/m2). Various models of turbulence have been tested. The results of the studies demonstrated a reasonable good agreement between CFD predictions and experimental data on the heat transfer coefficient and internal-wall temperature with use the SST turbulence model. Comparison of the CFD simulation data, which correspond to the presence or absence of the buoyancy forces, was performed. The regularities of the influence of these forces on the damping of turbulent transport, the deformation of the radial profiles of velocity and temperature along the channel length, the reduction of the heat transfer coefficients, etc. were studied. The features of the motion of the pseudo-phase transition within various conditions are presented.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"65 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":"132630766","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}
P. Wrigley, P. Wood, Paul Stewart, R. Hall, D. Robertson
The UK Small Modular Reactor (UKSMR) programme has been established to develop an SMR for the UK energy market. Developing an SMR is a multi-disciplinary technical challenge, involving nuclear physics, electrical, mechanical, design, management, safety, testing to name but a few. In 2016 Upadhyay & Jain performed a literature review on modularity in Nuclear Power. They concluded that although modularisation has been utilised in nuclear to reduce costs, more work needs to be done to “create effective modules”. Hohmann et al also concluded the same for defining modules in the chemical process plant industry. The aim of this paper is to further define modules with a particular focus on an SMR for the UK market, the UKSMR. The methods highlighted may be relevant and applied to other international SMR designs or other types of plant. An overview and examination of modularisation work in nuclear to date is provided. The different configurations are defined for the Nuclear Steam Supply System (NSSS) in primary circuits and then for Balance of Plant (BOP) modules. A top level design process has been defined to aid in the understanding of design choices for current reactors and to further assist designing balance of plant modules. The paper then highlights areas for additional research that may further support module design and definition.
{"title":"Design for Plant Modularisation: Nuclear and SMR","authors":"P. Wrigley, P. Wood, Paul Stewart, R. Hall, D. Robertson","doi":"10.1115/ICONE26-81760","DOIUrl":"https://doi.org/10.1115/ICONE26-81760","url":null,"abstract":"The UK Small Modular Reactor (UKSMR) programme has been established to develop an SMR for the UK energy market. Developing an SMR is a multi-disciplinary technical challenge, involving nuclear physics, electrical, mechanical, design, management, safety, testing to name but a few.\u0000 In 2016 Upadhyay & Jain performed a literature review on modularity in Nuclear Power. They concluded that although modularisation has been utilised in nuclear to reduce costs, more work needs to be done to “create effective modules”. Hohmann et al also concluded the same for defining modules in the chemical process plant industry.\u0000 The aim of this paper is to further define modules with a particular focus on an SMR for the UK market, the UKSMR. The methods highlighted may be relevant and applied to other international SMR designs or other types of plant.\u0000 An overview and examination of modularisation work in nuclear to date is provided. The different configurations are defined for the Nuclear Steam Supply System (NSSS) in primary circuits and then for Balance of Plant (BOP) modules. A top level design process has been defined to aid in the understanding of design choices for current reactors and to further assist designing balance of plant modules.\u0000 The paper then highlights areas for additional research that may further support module design and definition.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"68 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":"117337884","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}
T. Feng, W. Tian, P. Song, Jun Wang, Mingjun Wang, G. Su, S. Qiu
Core degradation experiment will be conducted in Fuel Rod Melt Progression apparatus (FROMA) to investigate the distribution of mass and energy in designed fuel rod during a core degradation process. Numerical research on the core degradation experiment is mainly to evaluate the designed parameters of FROMA. The pre-numerical study was conducted using the widely accepted severe accident analysis software MELCOR. The fuel rods used in the experiment consisted of real reactor fuel elements, and the designed fuel rods were electrically heated with internal center tungsten-rhenium rod to a very high temperature. Numerical analysis model is described, and the input parameters are in accord with experimental conditions. In the transient period of core degradation, the production hydrogen and the distribution of mass and temperature are obtained. All the predicted MELCOR results will be compared with the experimental measurements.
{"title":"Numerical Research on Fuel Rod Progression During Core Degradation Process Using MELCOR","authors":"T. Feng, W. Tian, P. Song, Jun Wang, Mingjun Wang, G. Su, S. Qiu","doi":"10.1115/ICONE26-81738","DOIUrl":"https://doi.org/10.1115/ICONE26-81738","url":null,"abstract":"Core degradation experiment will be conducted in Fuel Rod Melt Progression apparatus (FROMA) to investigate the distribution of mass and energy in designed fuel rod during a core degradation process. Numerical research on the core degradation experiment is mainly to evaluate the designed parameters of FROMA. The pre-numerical study was conducted using the widely accepted severe accident analysis software MELCOR. The fuel rods used in the experiment consisted of real reactor fuel elements, and the designed fuel rods were electrically heated with internal center tungsten-rhenium rod to a very high temperature. Numerical analysis model is described, and the input parameters are in accord with experimental conditions. In the transient period of core degradation, the production hydrogen and the distribution of mass and temperature are obtained. All the predicted MELCOR results will be compared with the experimental measurements.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"9 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":"123007868","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 Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.
{"title":"Heat Transfer and Fluid Flow Characteristics of One Side Heated Vertical Rectangular Channel Applied As Vessel Cooling System of VHTR","authors":"Kenta Fujikami, T. Takeda, S. Funatani","doi":"10.1115/ICONE26-81700","DOIUrl":"https://doi.org/10.1115/ICONE26-81700","url":null,"abstract":"A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"48 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":"114101368","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}
Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2]. In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.
{"title":"Operational Impacts and Consequences of Piping Component Failure: A Review of Operating Experience Data As Recorded in CODAP","authors":"Braedon Carr, B. Lydell, J. Riznic","doi":"10.1115/ICONE26-81001","DOIUrl":"https://doi.org/10.1115/ICONE26-81001","url":null,"abstract":"Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2].\u0000 In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 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":"121701025","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 vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs. Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle. Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N2 mixture, application of SuperCritical (SC)-CO2 cycles or combined cycles. The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO2 is provided.
{"title":"Research on Thermal Efficiencies of Various Power Cycles for GFRs and VHTRs","authors":"M. Mahdi, Roman Popov, I. Pioro","doi":"10.1115/ICONE26-81618","DOIUrl":"https://doi.org/10.1115/ICONE26-81618","url":null,"abstract":"The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs.\u0000 Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle.\u0000 Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N2 mixture, application of SuperCritical (SC)-CO2 cycles or combined cycles.\u0000 The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO2 is provided.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"111 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":"116576054","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}
I. K. Aji, Tokushima Tatsuya, M. Kinoshita, T. Okawa
Freeze valve technology is the main feature of safety system in the molten salt reactor. Freeze valve made from frozen salt located between reactor core and drain tank. The freeze valve will automatically melt and open on the accident condition, and respectively molten salt fuel will drain out from reactor core to the drain tanks. Melting time of frozen salt is important issues on this study, where draining process of the liquid fuel must be carried out immediately after the accident. Many factors affect to the opening time of freeze valve. On this experiment, describe a melting process of frozen salt which is affected by wall effect. HTS (high transfer salt) utilized as salt material, and a metal stick planted in the frozen HTS with a certain depth. The experimental process begins when the liquid HTS poured on the top of frozen HTS and ends when the metal stick detached from the frozen HTS. This experiments focus to analyze melting time of freeze valve which impacted by several parameters; material diameter which represent about material thickness in real situation, liquid temperature which represent about molten salt fuel, length of material which propose about the freeze valve thickness, and material difference which propose about thermal diffusivity effect. Results from the experiments will be utilized as a basic to developed mathematics and numerical analysis.
{"title":"An Experimental Study on Freeze Valve Performance in a Molten Salt Reactor","authors":"I. K. Aji, Tokushima Tatsuya, M. Kinoshita, T. Okawa","doi":"10.1115/ICONE26-81679","DOIUrl":"https://doi.org/10.1115/ICONE26-81679","url":null,"abstract":"Freeze valve technology is the main feature of safety system in the molten salt reactor. Freeze valve made from frozen salt located between reactor core and drain tank. The freeze valve will automatically melt and open on the accident condition, and respectively molten salt fuel will drain out from reactor core to the drain tanks. Melting time of frozen salt is important issues on this study, where draining process of the liquid fuel must be carried out immediately after the accident. Many factors affect to the opening time of freeze valve. On this experiment, describe a melting process of frozen salt which is affected by wall effect. HTS (high transfer salt) utilized as salt material, and a metal stick planted in the frozen HTS with a certain depth. The experimental process begins when the liquid HTS poured on the top of frozen HTS and ends when the metal stick detached from the frozen HTS. This experiments focus to analyze melting time of freeze valve which impacted by several parameters; material diameter which represent about material thickness in real situation, liquid temperature which represent about molten salt fuel, length of material which propose about the freeze valve thickness, and material difference which propose about thermal diffusivity effect. Results from the experiments will be utilized as a basic to developed mathematics and numerical analysis.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 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":"122736551","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}
Fan Zhichun, L. Mingze, Chen Feng, Huang Zhiyong, Yan He
Failure of sealing materials is a principal cause to the leakage of electrical penetration assemblies (EPA) in nuclear plants, and the essence can be attributed to small deformations or defects taking place in the sealing materials (glass, epoxy, etc.) as a result of harsh environment influence (high temperature, pressure and ionizing radiation), which leads to leakage when the stress/strain exceeds the threshold value. Metal-to-glass sealing EPA has unique advantages of higher temperature and pressure endurance than organic material sealing EPA, and has been applied in the pressure vessel of High-Temperature Reactor Pebble-bed Modules (HTR-PM) at the Shidao Bay Nuclear Power Plant in China. To achieve on-line state monitoring, we proposed a new method to monitor the stress in the sealing glass by optical fiber sensing technique. Our research showed that the stress in sealing glass could be measured via embedding an optical fiber Bragg grating (FBG) sensor in glass. Optical fiber sensing technique has been widely used for stress measurement in many fields, however applications in metal-to-glass sealing EPA have not been reported in the literature yet. Taking advantage of the small size of a fiber sensor, the embedding of fiber will not affect the sealing structure. And taking advantage of the similar chemical content, fiber and glass can be fused together easily without affecting insulation. In this paper, a brief review on applications of FBG in nuclear facilities was present. The model of FBG embedded EPA was built based on finite element method. Sensitivity analysis about the impact of environment parameters including temperature and pressure on stress had been studied numerically. And the theoretical Bragg wavelength shift of the embedded sensor was derived from the strain/stress distribution. Experiments had been carried out in some main aspects, including pressure and thermal test, from which the relationship between environment parameters and Bragg wavelength shifts was obtained. This research makes an initial attempt for realizing an on-line real-time long-term state monitoring and sets a base for the life cycle diagnostics of EPA in nuclear reactors.
{"title":"Stress Monitoring of Sealing Materials in Electrical Penetration Assemblies","authors":"Fan Zhichun, L. Mingze, Chen Feng, Huang Zhiyong, Yan He","doi":"10.1115/ICONE26-82165","DOIUrl":"https://doi.org/10.1115/ICONE26-82165","url":null,"abstract":"Failure of sealing materials is a principal cause to the leakage of electrical penetration assemblies (EPA) in nuclear plants, and the essence can be attributed to small deformations or defects taking place in the sealing materials (glass, epoxy, etc.) as a result of harsh environment influence (high temperature, pressure and ionizing radiation), which leads to leakage when the stress/strain exceeds the threshold value. Metal-to-glass sealing EPA has unique advantages of higher temperature and pressure endurance than organic material sealing EPA, and has been applied in the pressure vessel of High-Temperature Reactor Pebble-bed Modules (HTR-PM) at the Shidao Bay Nuclear Power Plant in China. To achieve on-line state monitoring, we proposed a new method to monitor the stress in the sealing glass by optical fiber sensing technique. Our research showed that the stress in sealing glass could be measured via embedding an optical fiber Bragg grating (FBG) sensor in glass. Optical fiber sensing technique has been widely used for stress measurement in many fields, however applications in metal-to-glass sealing EPA have not been reported in the literature yet. Taking advantage of the small size of a fiber sensor, the embedding of fiber will not affect the sealing structure. And taking advantage of the similar chemical content, fiber and glass can be fused together easily without affecting insulation. In this paper, a brief review on applications of FBG in nuclear facilities was present. The model of FBG embedded EPA was built based on finite element method. Sensitivity analysis about the impact of environment parameters including temperature and pressure on stress had been studied numerically. And the theoretical Bragg wavelength shift of the embedded sensor was derived from the strain/stress distribution. Experiments had been carried out in some main aspects, including pressure and thermal test, from which the relationship between environment parameters and Bragg wavelength shifts was obtained. This research makes an initial attempt for realizing an on-line real-time long-term state monitoring and sets a base for the life cycle diagnostics of EPA in nuclear reactors.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"120 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":"133918089","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}
Lorenzo Basili, R. L. Frano, M. Olcese, I. Sekachev, D. Aquaro
The aim of the paper is to investigate the thermal conditions (temperature distribution, heat losses) in the support system of the Vapour Suppression Tank (VST) of the Vacuum Vessel Pressure Suppression System (VVPSS), a safety important system of ITER fusion reactor, protecting the Vacuum Vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and mounted as two stacked assemblies. The study focuses on the optimization of the design of the thermal insulation of the bottom part of the VST, connecting each two-tank stack to the basement, and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary. A Computational Fluid Dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady state and transient loads following accidental conditions. The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.
{"title":"CFD Thermal Analysis of ITER Pressure Suppression Tanks","authors":"Lorenzo Basili, R. L. Frano, M. Olcese, I. Sekachev, D. Aquaro","doi":"10.1115/ICONE26-82550","DOIUrl":"https://doi.org/10.1115/ICONE26-82550","url":null,"abstract":"The aim of the paper is to investigate the thermal conditions (temperature distribution, heat losses) in the support system of the Vapour Suppression Tank (VST) of the Vacuum Vessel Pressure Suppression System (VVPSS), a safety important system of ITER fusion reactor, protecting the Vacuum Vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and mounted as two stacked assemblies.\u0000 The study focuses on the optimization of the design of the thermal insulation of the bottom part of the VST, connecting each two-tank stack to the basement, and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary.\u0000 A Computational Fluid Dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady state and transient loads following accidental conditions.\u0000 The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"193 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":"132199199","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}
Hsun-Chia Lin, S. Zhang, S. Shi, Xiaodong Sun, Richard M. Christensen
The Advanced High Temperature Reactor (AHTR) is a fluoride-salt-cooled high-temperature reactor (FHR) design concept that is currently being developed at Oak Ridge National Laboratory for efficient production of electricity with improved safety features. Transient analyses of different scenarios are critical to demonstrate the safety of the AHTR design. An AHTR reactor model has been developed using RELAP5/SCDAPSIM/MOD 4.0. Thermodynamic and transport properties of three molten fluoride salts, namely FLiBe, FLiNaK, and KF-ZrF4, have been implemented into the RELAP5 code. The AHTR RELAP5 model consists of a reactor core, an upper plenum, a lower plenum, three primary loops, and three Direct Reactor Auxiliary Cooling Systems (DRACS) loops. DRACS Heat Exchangers (DHX) and Natural Draft Heat Exchangers (NDHX) are important components of DRACS and provide coupling between the primary loops and DRACS loops, and DRACS loops and air chimneys, respectively. Single-wall fluted tube heat exchanger designs have been proposed for the DHX and the NDHX to improve heat transfer performance in the two heat exchangers, and heat transfer correlations for fluted tubes have also been implemented into the RELAP5 code. In this study, steady-state reactor normal operation and two transient scenarios are analyzed with the RELAP5 AHTR model. Based on a thermal hydraulics Phenomena Identification Ranking Table (PIRT) exercise, loss of forced circulation (LOFC) and loss of multiple DRACS loops are selected as the two transients for analysis. During transients, the decay heat is removed by the ambient air, fully relying on natural circulation/convection. The results of both transient scenarios show sufficient decay heat removal capabilities of DRACS with the proposed design.
{"title":"Transient Modeling of Advanced High Temperature Reactor (AHTR) in RELAP5/SCDAPSIM/MOD 4.0","authors":"Hsun-Chia Lin, S. Zhang, S. Shi, Xiaodong Sun, Richard M. Christensen","doi":"10.1115/ICONE26-81874","DOIUrl":"https://doi.org/10.1115/ICONE26-81874","url":null,"abstract":"The Advanced High Temperature Reactor (AHTR) is a fluoride-salt-cooled high-temperature reactor (FHR) design concept that is currently being developed at Oak Ridge National Laboratory for efficient production of electricity with improved safety features. Transient analyses of different scenarios are critical to demonstrate the safety of the AHTR design. An AHTR reactor model has been developed using RELAP5/SCDAPSIM/MOD 4.0. Thermodynamic and transport properties of three molten fluoride salts, namely FLiBe, FLiNaK, and KF-ZrF4, have been implemented into the RELAP5 code. The AHTR RELAP5 model consists of a reactor core, an upper plenum, a lower plenum, three primary loops, and three Direct Reactor Auxiliary Cooling Systems (DRACS) loops. DRACS Heat Exchangers (DHX) and Natural Draft Heat Exchangers (NDHX) are important components of DRACS and provide coupling between the primary loops and DRACS loops, and DRACS loops and air chimneys, respectively. Single-wall fluted tube heat exchanger designs have been proposed for the DHX and the NDHX to improve heat transfer performance in the two heat exchangers, and heat transfer correlations for fluted tubes have also been implemented into the RELAP5 code.\u0000 In this study, steady-state reactor normal operation and two transient scenarios are analyzed with the RELAP5 AHTR model. Based on a thermal hydraulics Phenomena Identification Ranking Table (PIRT) exercise, loss of forced circulation (LOFC) and loss of multiple DRACS loops are selected as the two transients for analysis. During transients, the decay heat is removed by the ambient air, fully relying on natural circulation/convection. The results of both transient scenarios show sufficient decay heat removal capabilities of DRACS with the proposed design.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"21 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":"132337232","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：