Pushpanjay K. Singh, R. Kothari, S. Sahu, P. K. Upadhyay, Shashwat Singh
Synthetic jet potentially useful in electronics cooling is investigated both numerically and experimentally. In the present study, a confined three dimensional synthetic jet with sinusoidal moving wall is considered. Computations are carried out using the FLUENT software with the coupled user defined function describing the diaphragm movement. In this study the effect of various geometrical parameters influencing the flow field and heat transfer are investigated. The effects of change in orifice geometry (circular, square and rectangular), orifice aspect ratio, and jet-to-plate distance are studied for a given hydraulic diameter. The heat transfer results obtained from the synthetic jet is compared with the continuous jet. An electromagnetic actuator is used as an oscillating diaphragm for the generation of synthetic jet. A stainless steel foil with 0.05 mm thickness is used as the test specimen. The surface temperature of the test specimen is measured by using a thermal imaging technique during synthetic jet impingement and a constant temperature anemometer has been employed for velocity measurement. Tests are carried out for Reynolds number of 5448, varied range of jet-to-plate distance (1–18). The maximum value of the heat transfer coefficient is found to be 16 times more than the heat transfer coefficient for natural convection.
{"title":"Experimental and Numerical Investigation of Thermal Performance of Synthetic Jet Impingement","authors":"Pushpanjay K. Singh, R. Kothari, S. Sahu, P. K. Upadhyay, Shashwat Singh","doi":"10.1115/icone2020-16775","DOIUrl":"https://doi.org/10.1115/icone2020-16775","url":null,"abstract":"\u0000 Synthetic jet potentially useful in electronics cooling is investigated both numerically and experimentally. In the present study, a confined three dimensional synthetic jet with sinusoidal moving wall is considered. Computations are carried out using the FLUENT software with the coupled user defined function describing the diaphragm movement. In this study the effect of various geometrical parameters influencing the flow field and heat transfer are investigated. The effects of change in orifice geometry (circular, square and rectangular), orifice aspect ratio, and jet-to-plate distance are studied for a given hydraulic diameter. The heat transfer results obtained from the synthetic jet is compared with the continuous jet. An electromagnetic actuator is used as an oscillating diaphragm for the generation of synthetic jet. A stainless steel foil with 0.05 mm thickness is used as the test specimen. The surface temperature of the test specimen is measured by using a thermal imaging technique during synthetic jet impingement and a constant temperature anemometer has been employed for velocity measurement. Tests are carried out for Reynolds number of 5448, varied range of jet-to-plate distance (1–18). The maximum value of the heat transfer coefficient is found to be 16 times more than the heat transfer coefficient for natural convection.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85834232","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
S. Qiao, W. Zhong, Hao Sijia, Peiyao Qi, Sichao Tan
The present study investigates the air-water two-phase flow across a 90-degree vertical-upward elbow with the computational fluid dynamics (CFD) simulation. The Eulerian-Eulerian two-fluid model and the Multi Size Group (MUSIG) model are used to predict the development of the detailed interfacial structures between the two phases. The axial development of the void fraction and the interfacial area concentration are investigated and benchmarked with the experimental data measured using the four-sensor conductivity probe. It is concluded that CFD simulation can predict the characteristics distributions of void fraction and interfacial area concentration and their development downstream of the elbow. The double-peaked void fraction distribution is found to be caused by the secondary flow induced by the elbow. The liquid phase on the outer curvature moves to the inner curvature and forms a double counter rotating vortex, entraining the bubbles to form a double-peaked distribution. The elbow effects become dissipated between 33 and 63 hydraulic diameters. The simulation results of liquid-phase and gas-phase parameters can be used to develop the theoretical two-phase flow models for the elbow region.
{"title":"Numerical Simulation of the Air-Water Two-Phase Flow Across a 90-Degree Vertical-Upward Elbow","authors":"S. Qiao, W. Zhong, Hao Sijia, Peiyao Qi, Sichao Tan","doi":"10.1115/icone2020-16836","DOIUrl":"https://doi.org/10.1115/icone2020-16836","url":null,"abstract":"\u0000 The present study investigates the air-water two-phase flow across a 90-degree vertical-upward elbow with the computational fluid dynamics (CFD) simulation. The Eulerian-Eulerian two-fluid model and the Multi Size Group (MUSIG) model are used to predict the development of the detailed interfacial structures between the two phases. The axial development of the void fraction and the interfacial area concentration are investigated and benchmarked with the experimental data measured using the four-sensor conductivity probe. It is concluded that CFD simulation can predict the characteristics distributions of void fraction and interfacial area concentration and their development downstream of the elbow. The double-peaked void fraction distribution is found to be caused by the secondary flow induced by the elbow. The liquid phase on the outer curvature moves to the inner curvature and forms a double counter rotating vortex, entraining the bubbles to form a double-peaked distribution. The elbow effects become dissipated between 33 and 63 hydraulic diameters. The simulation results of liquid-phase and gas-phase parameters can be used to develop the theoretical two-phase flow models for the elbow region.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87358536","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}
Flow regimes at water pressures above the critical pressure are characterized as supercritical flow. Supercritical flows have no phase change. The heat transfer from the wall to the fluid is single phase (there is no boiling or condensation). Experimental data indicate that for conditions that involve supercritical single-phase heat transfer, the Dittus-Boelter heat transfer correlation can be in significant error. A pseudo-critical temperature can be defined as a function of pressure for pressures that exceed the supercritical pressure. The pseudo-critical temperature is defined for heat transfer purposes as the temperature at which the specific heat peaks as the pressure is held constant. There is significant variation in fluid properties across the heat transfer boundary layer at temperatures near the pseudo-critical temperature. The large variation in properties is the reason for the failure of the Dittus-Boelter heat transfer correlation. Comparisons to experimental data indicate that the Mokry heat transfer correlation is a significant improvement over the Dittus-Boelter heat transfer correlation for single phase supercritical heat transfer. The Mokry correlation was chosen to be included into TRACE.
{"title":"Heat Transfer for Supercritical Flow With TRACE","authors":"J. Spore, G. Roth","doi":"10.1115/icone2020-16753","DOIUrl":"https://doi.org/10.1115/icone2020-16753","url":null,"abstract":"\u0000 Flow regimes at water pressures above the critical pressure are characterized as supercritical flow. Supercritical flows have no phase change. The heat transfer from the wall to the fluid is single phase (there is no boiling or condensation). Experimental data indicate that for conditions that involve supercritical single-phase heat transfer, the Dittus-Boelter heat transfer correlation can be in significant error.\u0000 A pseudo-critical temperature can be defined as a function of pressure for pressures that exceed the supercritical pressure. The pseudo-critical temperature is defined for heat transfer purposes as the temperature at which the specific heat peaks as the pressure is held constant. There is significant variation in fluid properties across the heat transfer boundary layer at temperatures near the pseudo-critical temperature. The large variation in properties is the reason for the failure of the Dittus-Boelter heat transfer correlation.\u0000 Comparisons to experimental data indicate that the Mokry heat transfer correlation is a significant improvement over the Dittus-Boelter heat transfer correlation for single phase supercritical heat transfer. The Mokry correlation was chosen to be included into TRACE.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91196799","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}
SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.
{"title":"Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions","authors":"Chunyu Yin","doi":"10.1115/icone2020-16793","DOIUrl":"https://doi.org/10.1115/icone2020-16793","url":null,"abstract":"\u0000 SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90661782","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}
One sees eerie similarities here in Canada to the cozy relationship between regulator and utilities in ‘pre-Fukushima’ Japan. Such ties are hardly limited to Canada though. The chronic degradation of real commitments to continued improvements in reactor safety systems and a decline in overall safety culture that discourages critical design reviews and willfully ignores well justified, safety critical hardware upgrades, has created alarming conditions that are likely inching us towards another nuclear disaster. Operating CANDU reactors are now close to being obsolete but have barely seen any substantive severe accident related risk reduction upgrades nine years after Fukushima, hoopla in Canada around some minor improvements and premature closure of even otherwise sparse and what were really weak regulatory ‘Fukushima Action Items’ notwithstanding.
{"title":"Unmet Challenges to Successfully Mitigating Severe Accidents in Multi Unit CANDU Reactors","authors":"S. Nijhawan","doi":"10.1115/icone2020-16517","DOIUrl":"https://doi.org/10.1115/icone2020-16517","url":null,"abstract":"\u0000 One sees eerie similarities here in Canada to the cozy relationship between regulator and utilities in ‘pre-Fukushima’ Japan. Such ties are hardly limited to Canada though. The chronic degradation of real commitments to continued improvements in reactor safety systems and a decline in overall safety culture that discourages critical design reviews and willfully ignores well justified, safety critical hardware upgrades, has created alarming conditions that are likely inching us towards another nuclear disaster. Operating CANDU reactors are now close to being obsolete but have barely seen any substantive severe accident related risk reduction upgrades nine years after Fukushima, hoopla in Canada around some minor improvements and premature closure of even otherwise sparse and what were really weak regulatory ‘Fukushima Action Items’ notwithstanding.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82909324","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Takashi Mawatari, Yamamoto Yasushi, O. Shibasaki, T. Hara, Yusuke Horayama, J. Takagi
It is important to avoid a risk of stress corrosion cracking (SCC) of structure materials in nuclear power plants. One of the methods for preventing SCC is titanium-dioxide (TiO2) injection. TiO2 is a photo-catalyst, and injected TiO2 deposits on the surfaces of structure materials, and the deposited TiO2 can decrease the electrochemical corrosion potential (ECP) by irradiation of Cherenkov-light in the reactor. Generally, the ECP is an important indicator about susceptibility to SCC, and ECP-lowering makes a risk of SCC lower. Therefore, initiation and propagation of SCC will be mitigated by the TiO2 injection. The evaluation model for corrosive environment had been already developed. That “conventional” model was consisted of some analysis such as water radiolysis, ECP and others. However, the model did not consider an effect of TiO2 deposition and was one-dimensional simple model. Therefore, to evaluate ECP distribution in a reactor after TiO2 injection, the “newly developed” model was build: analyses integration, high-resolution and three-dimensional model. In this paper, the TiO2 deposition and water radiolysis analyses integrated into flow analysis were contained. Both the deposition and the radiolysis depend on flow in the RPV, so that these analyses were built into the computational fluid dynamics (CFD) analysis model for CFD code. The target plant type for the model was 1100 MWe-class BWR-5. The flow analysis was carried out at steady-state, assuming the TiO2 injection during the rated operation of the plant. Then the TiO2 deposition and the radiolysis analyses were conducted by transient calculations using the result of flow, and the three-dimensional and high resolution results were obtained.
{"title":"Evaluations of TiO2 Deposition on Structure Surfaces and Water Radiolysis for the Corrosive Environment in a Reactor Pressure Vessel","authors":"Takashi Mawatari, Yamamoto Yasushi, O. Shibasaki, T. Hara, Yusuke Horayama, J. Takagi","doi":"10.1115/icone2020-16405","DOIUrl":"https://doi.org/10.1115/icone2020-16405","url":null,"abstract":"\u0000 It is important to avoid a risk of stress corrosion cracking (SCC) of structure materials in nuclear power plants. One of the methods for preventing SCC is titanium-dioxide (TiO2) injection. TiO2 is a photo-catalyst, and injected TiO2 deposits on the surfaces of structure materials, and the deposited TiO2 can decrease the electrochemical corrosion potential (ECP) by irradiation of Cherenkov-light in the reactor. Generally, the ECP is an important indicator about susceptibility to SCC, and ECP-lowering makes a risk of SCC lower. Therefore, initiation and propagation of SCC will be mitigated by the TiO2 injection.\u0000 The evaluation model for corrosive environment had been already developed. That “conventional” model was consisted of some analysis such as water radiolysis, ECP and others. However, the model did not consider an effect of TiO2 deposition and was one-dimensional simple model. Therefore, to evaluate ECP distribution in a reactor after TiO2 injection, the “newly developed” model was build: analyses integration, high-resolution and three-dimensional model.\u0000 In this paper, the TiO2 deposition and water radiolysis analyses integrated into flow analysis were contained. Both the deposition and the radiolysis depend on flow in the RPV, so that these analyses were built into the computational fluid dynamics (CFD) analysis model for CFD code. The target plant type for the model was 1100 MWe-class BWR-5.\u0000 The flow analysis was carried out at steady-state, assuming the TiO2 injection during the rated operation of the plant. Then the TiO2 deposition and the radiolysis analyses were conducted by transient calculations using the result of flow, and the three-dimensional and high resolution results were obtained.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84161149","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 centrifugal pump is the key equipment for power conversion in the CAP1400 power plant. According to the equipment layout requirements in the nuclear power plant process system design manual, the pipe section at the entrance should be a straight pipe with a length at least five times the pipe diameter. However, in actual layout, due to factors such as plant space or modularity, the layout of the inlet pipes of some centrifugal pumps cannot meet this requirement, resulting in non-uniform incoming flow at the pump inlet, which affects the safe and stable operation of the pump and system. In this paper, two different structures of rectifiers are designed and numerically simulated for the elbow part of the inlet pipe section of the SFS system centrifugal pump in the CAP1400 power plant. Firstly, the variation of flow uniformity of the pump inlet pipe at different inlet speeds is studied. Then, the influence of different structure rectifiers on the uniformity of flow at the pump inlet under the same working conditions is studied. The results show that the addition of a rectifier in the elbow can effectively improve the flow uniformity of the pump inlet section and reduce the effect of uneven inlet flow on the pump performance.
{"title":"Numerical Simulation of Structural Optimization for Inlet Pipe of Centrifugal Pump in CAP1400 Power Plant Auxiliary System","authors":"Ting Qu, Liang Zhang","doi":"10.1115/icone2020-16673","DOIUrl":"https://doi.org/10.1115/icone2020-16673","url":null,"abstract":"\u0000 The centrifugal pump is the key equipment for power conversion in the CAP1400 power plant. According to the equipment layout requirements in the nuclear power plant process system design manual, the pipe section at the entrance should be a straight pipe with a length at least five times the pipe diameter. However, in actual layout, due to factors such as plant space or modularity, the layout of the inlet pipes of some centrifugal pumps cannot meet this requirement, resulting in non-uniform incoming flow at the pump inlet, which affects the safe and stable operation of the pump and system. In this paper, two different structures of rectifiers are designed and numerically simulated for the elbow part of the inlet pipe section of the SFS system centrifugal pump in the CAP1400 power plant. Firstly, the variation of flow uniformity of the pump inlet pipe at different inlet speeds is studied. Then, the influence of different structure rectifiers on the uniformity of flow at the pump inlet under the same working conditions is studied. The results show that the addition of a rectifier in the elbow can effectively improve the flow uniformity of the pump inlet section and reduce the effect of uneven inlet flow on the pump performance.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81847977","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 order to rationalize maintenance for nuclear power plants, it is necessary to develop optimize maintenance plan by considering characteristics of each plant. In sodium-cooled fast reactor, there are constraints on inspections due to the specialty of handling sodium equipment, that is one of the important points when considering rationalization of maintenance. To solve this problem, we proposed a basic concept of maintenance optimization scheme that is a design support tool in order to develop maintenance strategy, based on “system based code (SBC)”. SBC is a concept to optimize the reliability of a nuclear power plant by consideration of all related technical requirements. “ASME Code Case N-875” and “ASME Boiler and Pressure Vessel Code, Section XI, Division 2 (RIM)” based on system based code were already developed as standards for inspection. One of the proposed scheme goals is to make a concrete way of necessary assessment method. Another is to provide several combinations of design and maintenance, and information for owner in order to choose the acceptable combination. In the beginning, we are working to develop the scheme that can be applied to sodium fast reactor as the main concept of next generation reactor. In this context, primary heat transfer system (PHTS) piping of fast reactor was evaluated by the scheme. This piping was chosen because it is major significant component and the inspection have constraint conditions that need preparation work. As a result, design candidate (e.g. single and double wall piping) and inspection candidate (e.g. ultrasonic testing and continues leakage monitoring) combinations along with benefit of each cases were provided.
为了使核电站的维修合理化，有必要根据各电厂的特点制定优化维修计划。在钠冷快堆中，由于处理钠设备的特殊性，检查受到限制，这是考虑维护合理化的重要问题之一。为了解决这一问题，提出了基于“基于系统的代码(system based code, SBC)”的维修优化方案的基本概念，即维修优化方案是制定维修策略的设计支持工具。SBC是一种通过考虑所有相关技术要求来优化核电站可靠性的概念。提出的方案目标之一是制定必要的评估方法的具体途径。另一种是提供多种设计与维护组合，并提供信息供业主选择可接受的组合。在开始阶段，我们正在努力开发可以应用于钠快堆的方案，作为下一代反应堆的主要概念。在此背景下，用该方案对快堆一次传热系统(PHTS)管道进行了评价。选择该管道是因为它是重要的部件，而且检查有约束条件，需要做准备工作。因此，提供了候选设计(例如单壁和双壁管道)和候选检查(例如超声波测试和持续泄漏监测)组合以及每种情况的优点。
{"title":"Proposal of Inspection Rationalization Method and Application for Sodium Cooled Fast Reactor","authors":"Yada Hiroki, Takaya Shigeru, Enuma Yasuhiro","doi":"10.1115/icone2020-16735","DOIUrl":"https://doi.org/10.1115/icone2020-16735","url":null,"abstract":"\u0000 In order to rationalize maintenance for nuclear power plants, it is necessary to develop optimize maintenance plan by considering characteristics of each plant. In sodium-cooled fast reactor, there are constraints on inspections due to the specialty of handling sodium equipment, that is one of the important points when considering rationalization of maintenance. To solve this problem, we proposed a basic concept of maintenance optimization scheme that is a design support tool in order to develop maintenance strategy, based on “system based code (SBC)”. SBC is a concept to optimize the reliability of a nuclear power plant by consideration of all related technical requirements. “ASME Code Case N-875” and “ASME Boiler and Pressure Vessel Code, Section XI, Division 2 (RIM)” based on system based code were already developed as standards for inspection. One of the proposed scheme goals is to make a concrete way of necessary assessment method. Another is to provide several combinations of design and maintenance, and information for owner in order to choose the acceptable combination. In the beginning, we are working to develop the scheme that can be applied to sodium fast reactor as the main concept of next generation reactor. In this context, primary heat transfer system (PHTS) piping of fast reactor was evaluated by the scheme. This piping was chosen because it is major significant component and the inspection have constraint conditions that need preparation work. As a result, design candidate (e.g. single and double wall piping) and inspection candidate (e.g. ultrasonic testing and continues leakage monitoring) combinations along with benefit of each cases were provided.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86130297","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. Nguyen, Won-tae Kim, Seungpyo Hong, Haein Lee, Ahram Lee
Acoustic-induced vibration in piping system and other devices leads to premature wear and failure. Especially, in nuclear power plants, very high velocity and temperature gas flows inside pipe systems. Moreover, if a frequency due to the vibration in the piping system is overlapped with a natural frequency of the stud pipe, the magnitude of the amplitude will be increased resulting in severe failure. For example, damage can be considered as flow-induced acoustic resonance at the branch pipes of the safety relief valve in the main steam lines. Specially, the pipe system not only has multi-branches but also includes the elbow that the resonance could occurs making pressure oscillation stronger than that of a single branch because of the interaction between the branches and the elbow. This study has investigated a Computational Fluid Dynamics (CFD) analysis methodology to predict and quantify the vortex shedding frequencies and the pressure pulsation magnitude in the dead-end pipe. The influence of the pressure fluctuation amplitude between each branch, number of branch, and elbow is also investigated.
{"title":"Feasibility Study for Multi-Branches and Elbow in Stub Pipe System due to Flow-Induced Vibration Using Computational Fluid Dynamics","authors":"K. Nguyen, Won-tae Kim, Seungpyo Hong, Haein Lee, Ahram Lee","doi":"10.1115/icone2020-16431","DOIUrl":"https://doi.org/10.1115/icone2020-16431","url":null,"abstract":"Acoustic-induced vibration in piping system and other devices leads to premature wear and failure. Especially, in nuclear power plants, very high velocity and temperature gas flows inside pipe systems. Moreover, if a frequency due to the vibration in the piping system is overlapped with a natural frequency of the stud pipe, the magnitude of the amplitude will be increased resulting in severe failure. For example, damage can be considered as flow-induced acoustic resonance at the branch pipes of the safety relief valve in the main steam lines. Specially, the pipe system not only has multi-branches but also includes the elbow that the resonance could occurs making pressure oscillation stronger than that of a single branch because of the interaction between the branches and the elbow. This study has investigated a Computational Fluid Dynamics (CFD) analysis methodology to predict and quantify the vortex shedding frequencies and the pressure pulsation magnitude in the dead-end pipe. The influence of the pressure fluctuation amplitude between each branch, number of branch, and elbow is also investigated.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89843831","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}
W. Dongwei, L. Mingxing, Wu Xiao, Y. Hao, Wu Zhiqiang
Offshore floating nuclear power plant (FNPP) is characterized by its small and mobility, which is not only able to provide safe and efficient electric energy to remote islands, but to the oil and gas platforms. The safety digital control system (DCS) cabinet, as a carrier for the electronic devices, plays a significant role in ensuring the normal operation of the nuclear power plant. To satisfy the requirements of cabinet used in the sea environment, such as well rigidity, shock load resistance, good seal and corrosion resistance, etc, more and more attention is focused on the cast aluminum cabinet. However, the cast aluminum structure may cause larger weight of cabinet, which inevitability affects the mobility of cabinet, and increases the carried load of ship as well. Therefore, seeking for an effective approach to design a light weight cast aluminum cabinet for the offshore FNPP is definitely necessary. In this work, a frame of cast aluminum cabinet with lightweight is obtained successfully via structure topology optimization design, it is found that the weight of the frame can be reduced to 50% after optimization iterations. Subsequently, the natural frequency of the optimized cast aluminum cabinet is calculated by using ABAQUS, it is seen that the first mode frequency of the frame is beyond 30 Hz, which can meet the basic stiffness requirement. Accordingly, dynamic design analysis method (DDAM) is performed to verify the ability of the optimized cast aluminum cabinet in resisting sudden shock load, and the shock response characteristics of the cabinet are determined. Numerical results support that the optimized frame of cabinet possesses good resistance to high level shock. However, for the assembled cast aluminum cabinet, the vertical shock circumstance turns out to be the most critical condition, high stress and deformation regions occurs at the bracket and column. Reinforcements are proposed to make the bracket stiffer in this shock loading condition.
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