Betavoltaics are direct conversion energy devices that are ideal for low, micropower and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radioisotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radioiso-tope. The source density can be increased by the selection of solid-state substrates. While solid-state substrates can be selected from simulations, the viability of the substrate to absorb tritium has to evaluated. The development of a hydrogen loading system was performed to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert method, where the initial pressure and volume is constant and pressure drop in the system is used to determine hydrogen uptake of a film substrate. The procedures of the hydrogen loading system are detailed. To test the procedures of the hydrogen loading system, old, palladium films were loaded. Results show uptake of hydrogen by the thin palladium films, as well as cycles of hydrogen absorption and desorption. Hydrogen loading of palladium was compared to a prior result and was shown to have similar results.
{"title":"Hydrogen Loading System for Thin Films for Betavoltaics","authors":"D. Cheu, T. Adams, S. Revankar","doi":"10.1115/icone29-93910","DOIUrl":"https://doi.org/10.1115/icone29-93910","url":null,"abstract":"\u0000 Betavoltaics are direct conversion energy devices that are ideal for low, micropower and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radioisotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radioiso-tope. The source density can be increased by the selection of solid-state substrates. While solid-state substrates can be selected from simulations, the viability of the substrate to absorb tritium has to evaluated. The development of a hydrogen loading system was performed to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert method, where the initial pressure and volume is constant and pressure drop in the system is used to determine hydrogen uptake of a film substrate. The procedures of the hydrogen loading system are detailed. To test the procedures of the hydrogen loading system, old, palladium films were loaded. Results show uptake of hydrogen by the thin palladium films, as well as cycles of hydrogen absorption and desorption. Hydrogen loading of palladium was compared to a prior result and was shown to have similar results.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"31 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129905261","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}
Based on the measured dynamic range of fast neutron injection rate, 4H-SiC PIN detector fast neutron detection efficiency and detector output current signal distribution characteristics, this paper designs and develops a current sensitive front-end readout circuit based on a high bandwidth transimpedance amplifier (ADI LTC6268-10), analyzes the main factors affecting the front-end circuit characteristics, and explores the main methods to improve the linear range of the front-end circuit. The main methods to improve the linear range of the front-end circuit are explored. Second, according to the front-end circuit output signal characteristics, based on high-bandwidth ADC (AD9680) and Zynq UltraScale+ MPSoC XCZU15EG development board, a digital data acquisition system (DAQ) was designed and developed. The preliminary simulation and experimental test results show that the designed detector electronics circuit has a good linear response under the conditions of input current pulse frequency of 20MHz, peak value of 10μA∼100μA and pulse width of 5ns. The research results provide relevant technical support for the establishment of 4H-SiC detector electronics system.
{"title":"Preliminary Design on the Wide Dynamic and Fast Response Electronic System for the 4H-SiC PIN Fast Neutron Detector","authors":"Ma Yong, Liu Shuhuan, Liu Shuangying","doi":"10.1115/icone29-92416","DOIUrl":"https://doi.org/10.1115/icone29-92416","url":null,"abstract":"\u0000 Based on the measured dynamic range of fast neutron injection rate, 4H-SiC PIN detector fast neutron detection efficiency and detector output current signal distribution characteristics, this paper designs and develops a current sensitive front-end readout circuit based on a high bandwidth transimpedance amplifier (ADI LTC6268-10), analyzes the main factors affecting the front-end circuit characteristics, and explores the main methods to improve the linear range of the front-end circuit. The main methods to improve the linear range of the front-end circuit are explored. Second, according to the front-end circuit output signal characteristics, based on high-bandwidth ADC (AD9680) and Zynq UltraScale+ MPSoC XCZU15EG development board, a digital data acquisition system (DAQ) was designed and developed. The preliminary simulation and experimental test results show that the designed detector electronics circuit has a good linear response under the conditions of input current pulse frequency of 20MHz, peak value of 10μA∼100μA and pulse width of 5ns. The research results provide relevant technical support for the establishment of 4H-SiC detector electronics system.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"81 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126674293","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}
Guanhui Xie, Runcheng Li, Tianyi Wei, Zhikang Lin, Dongyang Li, S. Tan
In case of core emergency cooling safety injection, the boron containing coolant injected externally and dilution water mass will be mixed in the pressure vessel. The uneven mixing of boron containing coolant and dilution water mass may lead to the return of the core to criticality. Therefore, it is of great significance to accurately measure the boron concentration distribution in the pressure vessel and study the mechanism affecting the mixing process for the safe operation of the reactor. In this paper, combined with the structural characteristics of HPR1000 pressure vessel, a visual experimental device is built through proportional modeling design, and the flow mixing process of diluted water mass, safety injection solution and coolant in the pressure vessel under double loop operation is obtained by using plane laser-induced fluorescence technology, the effects of different Reynolds numbers on the diffusion of boric acid were experimentally studied. The experimental results show that under the condition of double loop, the two inlets divide the circumferential direction into good arc and bad arc. The velocity distribution of fluid flowing through the bad arc is more uniform, while the vertical flow of fluid flowing through the good arc is uneven; The fluid flows downward rapidly after the intersection on the inferior arc side, and the intersection height is positively correlated with the cold pipe flow, while the fluid on both sides shows obvious mixing after the intersection on the superior arc side; Clear water mass cannot be avoided at the core inlet, but increasing the cold pipe flow and safety injection flow can reduce the time of clear water mass flowing through the core.
{"title":"Flow Mixing Characteristics in Double Loop Reactor Pressure Vessel Under Accident Conditions","authors":"Guanhui Xie, Runcheng Li, Tianyi Wei, Zhikang Lin, Dongyang Li, S. Tan","doi":"10.1115/icone29-89228","DOIUrl":"https://doi.org/10.1115/icone29-89228","url":null,"abstract":"\u0000 In case of core emergency cooling safety injection, the boron containing coolant injected externally and dilution water mass will be mixed in the pressure vessel. The uneven mixing of boron containing coolant and dilution water mass may lead to the return of the core to criticality. Therefore, it is of great significance to accurately measure the boron concentration distribution in the pressure vessel and study the mechanism affecting the mixing process for the safe operation of the reactor. In this paper, combined with the structural characteristics of HPR1000 pressure vessel, a visual experimental device is built through proportional modeling design, and the flow mixing process of diluted water mass, safety injection solution and coolant in the pressure vessel under double loop operation is obtained by using plane laser-induced fluorescence technology, the effects of different Reynolds numbers on the diffusion of boric acid were experimentally studied. The experimental results show that under the condition of double loop, the two inlets divide the circumferential direction into good arc and bad arc. The velocity distribution of fluid flowing through the bad arc is more uniform, while the vertical flow of fluid flowing through the good arc is uneven; The fluid flows downward rapidly after the intersection on the inferior arc side, and the intersection height is positively correlated with the cold pipe flow, while the fluid on both sides shows obvious mixing after the intersection on the superior arc side; Clear water mass cannot be avoided at the core inlet, but increasing the cold pipe flow and safety injection flow can reduce the time of clear water mass flowing through the core.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"33 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126931329","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}
Lintai Li, Yanmin Zhou, Yuan Li, Zhong-ning Sun, Yin Wang, Haifeng Gu, Song Ma
As a new filter material, metal fiber has good high-temperature resistance, moisture resistance, and radiation resistance. To explore the influence of the fiber’s primary structural parameters on the fiber’s performance, this paper experimentally studied the relationship between the flow resistance, filtration efficiency, and dust holding capacity of the metal fiber filter material and the thickness and solid volume fraction (SVF). Under the condition of ensuring uniform deposition of NaCl aerosol, we found that the deposition law of NaCl polydisperse aerosol is the same as that of monodisperse aerosol. The results show that for the same fiber diameter, increasing the fiber’s thickness and SVF will increase the flow resistance of the fiber under the same dust holding capacity; that is, reducing the dust holding capacity of the fiber. In the depth filtration, the efficiency increases rapidly due to the aerosol deposition in the fiber, while the efficiency changes gently when the filter cake is covered on the filtration surface. At the same time, for each high filtration efficiency fiber, the variation characteristics of resistance with dust holding capacity in the deep filtration stage are the same. After the filter cake is completely covered on the filter surface, the growth trend of resistance is the same, independent of the fiber’s structural parameters.
{"title":"Experimental Study on the Relationship Between Performance and Structural Parameters of Metal Fiber","authors":"Lintai Li, Yanmin Zhou, Yuan Li, Zhong-ning Sun, Yin Wang, Haifeng Gu, Song Ma","doi":"10.1115/icone29-91762","DOIUrl":"https://doi.org/10.1115/icone29-91762","url":null,"abstract":"\u0000 As a new filter material, metal fiber has good high-temperature resistance, moisture resistance, and radiation resistance. To explore the influence of the fiber’s primary structural parameters on the fiber’s performance, this paper experimentally studied the relationship between the flow resistance, filtration efficiency, and dust holding capacity of the metal fiber filter material and the thickness and solid volume fraction (SVF). Under the condition of ensuring uniform deposition of NaCl aerosol, we found that the deposition law of NaCl polydisperse aerosol is the same as that of monodisperse aerosol. The results show that for the same fiber diameter, increasing the fiber’s thickness and SVF will increase the flow resistance of the fiber under the same dust holding capacity; that is, reducing the dust holding capacity of the fiber. In the depth filtration, the efficiency increases rapidly due to the aerosol deposition in the fiber, while the efficiency changes gently when the filter cake is covered on the filtration surface. At the same time, for each high filtration efficiency fiber, the variation characteristics of resistance with dust holding capacity in the deep filtration stage are the same. After the filter cake is completely covered on the filter surface, the growth trend of resistance is the same, independent of the fiber’s structural parameters.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"89 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127193479","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}
Seigo Kai, Takeshi Moriya, Naruki Shoji, G. Endo, H. Takahashi, H. Kikura
On March 11, 2011, a serious accident occurred at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. due to the Great East Japan Earthquake and tsunami. Currently, nitrogen gas is used to fill the reactor containment vessel, pressure vessel, and plant piping to prevent a hydrogen explosion. Gas leak detection technology is needed to maintain and control the integrity of nitrogen gas containment and to ensure plant safety, as well as to identify the leak location in the event of a leak and understand the nature of the leak before plugging the leak location. However, because of the high radiation environment in the plant, it is not practical for a person to enter the plant to detect gas leakage. Therefore, there is a need for a leak detection method for gases that can be used in a radiation environment, and that is compact, measurable, and can be loaded by robots. In this study, we focused on acoustic gas leak detection as a leak detection technique. The characteristics of ultrasound generated by gas leakage were investigated for the development of gas leak detection technology and source localization system in nuclear reactors.
{"title":"Study for the Construction of Gas Leak Detection System in Nuclear Power Plants Using Ultrasound","authors":"Seigo Kai, Takeshi Moriya, Naruki Shoji, G. Endo, H. Takahashi, H. Kikura","doi":"10.1115/icone29-91695","DOIUrl":"https://doi.org/10.1115/icone29-91695","url":null,"abstract":"\u0000 On March 11, 2011, a serious accident occurred at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. due to the Great East Japan Earthquake and tsunami. Currently, nitrogen gas is used to fill the reactor containment vessel, pressure vessel, and plant piping to prevent a hydrogen explosion. Gas leak detection technology is needed to maintain and control the integrity of nitrogen gas containment and to ensure plant safety, as well as to identify the leak location in the event of a leak and understand the nature of the leak before plugging the leak location. However, because of the high radiation environment in the plant, it is not practical for a person to enter the plant to detect gas leakage. Therefore, there is a need for a leak detection method for gases that can be used in a radiation environment, and that is compact, measurable, and can be loaded by robots. In this study, we focused on acoustic gas leak detection as a leak detection technique. The characteristics of ultrasound generated by gas leakage were investigated for the development of gas leak detection technology and source localization system in nuclear reactors.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125682074","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}
Yidong Wu, Li Shi, Xinxin Wu, Xiaoxin Wang, Q. Xiao
Based on the design of the heat transfer tube bundles with square cross section of the steam generators in high temperature reactor-pebble bed modules (HTR-PM), a free-vibration experiment is conducted to examine flow-induced vibration (FIV) characteristic of two identical rounded square cylinders with r/D = 0.12 in tandem arrangements at spacing ratio L/D = 1.5∼5.5. One of the cylinders is two-dimensional, spring mounted, and allowed to vibrate in the cross-flow direction while the other is held stationary. Considering two cases that elastic cylinder located upstream or downstream, two mass-damping ratios m*ζ are chosen to investigate the effect of spacing between two cylinder. The vibration responses and the flow structure around two cylinders are studied, using laser displacement sensor and Particle image velocimetry. It is observed that only vortex-induced vibration (VIV) occurs when the elastic cylinder is located downstream. When elastic cylinder is located upstream, the full interaction between VIV and galloping of cylinder with lower m*ζ is invariable no matter what the spacing is. The spacing has noticeable effects on the vibration behavior of cylinder with higher m*ζ and changes the interaction between VIV and galloping. The flow structure of two cylinders indicates that the upstream shear layer reattaches on the surface of the downstream cylinder as L/D < 3.5, where the St drops with increasing L/D. There is gap vortices between two cylinders at L/D > 3.5 where the St rises and is close to the value of single cylinder with a larger spacing.
{"title":"Flow-Induced Vibration of Two Square Cylinders With Rounded Corners in a Tandem Arrangement","authors":"Yidong Wu, Li Shi, Xinxin Wu, Xiaoxin Wang, Q. Xiao","doi":"10.1115/icone29-90995","DOIUrl":"https://doi.org/10.1115/icone29-90995","url":null,"abstract":"\u0000 Based on the design of the heat transfer tube bundles with square cross section of the steam generators in high temperature reactor-pebble bed modules (HTR-PM), a free-vibration experiment is conducted to examine flow-induced vibration (FIV) characteristic of two identical rounded square cylinders with r/D = 0.12 in tandem arrangements at spacing ratio L/D = 1.5∼5.5. One of the cylinders is two-dimensional, spring mounted, and allowed to vibrate in the cross-flow direction while the other is held stationary. Considering two cases that elastic cylinder located upstream or downstream, two mass-damping ratios m*ζ are chosen to investigate the effect of spacing between two cylinder. The vibration responses and the flow structure around two cylinders are studied, using laser displacement sensor and Particle image velocimetry. It is observed that only vortex-induced vibration (VIV) occurs when the elastic cylinder is located downstream. When elastic cylinder is located upstream, the full interaction between VIV and galloping of cylinder with lower m*ζ is invariable no matter what the spacing is. The spacing has noticeable effects on the vibration behavior of cylinder with higher m*ζ and changes the interaction between VIV and galloping. The flow structure of two cylinders indicates that the upstream shear layer reattaches on the surface of the downstream cylinder as L/D < 3.5, where the St drops with increasing L/D. There is gap vortices between two cylinders at L/D > 3.5 where the St rises and is close to the value of single cylinder with a larger spacing.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"99 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127650342","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 the case of a severe accident, such as a loss of power supply, it is necessary to keep the nuclear power plant in a safe condition. Therefore, steam-water injectors (SI) have been attracting attention as a device that can cool the plant without power supply in the event of a severe accident. In a SI, the steam condenses by direct contact between steam and water interface. As the steam condenses, negative pressure is generated, and a force that attracts water is generated. The SI can process a large amount of cooling water with this force. It can also function as a reactor condenser for nuclear reactors due to its high condensation performance. There are some characteristics of SI that the discharge pressure is expected to exceed the inlet pressure after being accelerated at the throat, and it has a very simple structure with converging-diverging part, which is expected to reduce the cost of installation and maintenance. Previous studies have indicated that GROLMES et al. (1968) inferred the flow in the reduced test section by measuring the void fraction and MIWA et al. (2018) suggested a relationship between internal flow and the operating range of SI, referring to the results inferred by GROLMES et al. (1968). Although these existing studies have suggested some relationship between the internal flow and the operating range, the relationship between the stability of the water jet and the behavior of the SI has not been clarified. Therefore, the aim of this study is to clarify the effect of water jet behavior on the operating range of SI. The taper angle of the water jet outlet was varied and the relationship between the water jet behavior and the operating range was investigated. In this study, experiments were conducted using two types of taper angles at the water jet outlet: 20 and 7 degrees. As a result of detailed observation inside the SI, it was confirmed that the water jet with the taper angle of 7 degrees was less spreading than that with the taper angle of 20 degrees. In the operating range of this experiment, no significant difference was observed when the taper angle of the water jet was changed. However, it is suggested that the water jet with a certain degree of dispersion might be better in terms of heat exchange due to large surface area. These results suggest that the stability of the water jet is a necessary but not a sufficient condition for operation.
在发生严重事故（如失去电力供应）的情况下，有必要使核电站处于安全状态。因此，蒸汽-水喷射器（SI）作为一种能在发生严重事故时在没有电力供应的情况下冷却核电站的装置，一直备受关注。在蒸汽-水喷射器中，蒸汽通过与水界面的直接接触而冷凝。蒸汽冷凝时会产生负压，并产生吸引水的力。利用这种力量，SI 可以处理大量冷却水。由于冷凝性能高，它还可用作核反应堆的冷凝器。SI 的一些特点是，在喉部加速后，排出压力有望超过入口压力，而且它的汇流-分流部分结构非常简单，有望降低安装和维护成本。以往的研究表明，GROLMES 等人（1968 年）通过测量空隙率推断出缩小试验段的流量，MIWA 等人（2018 年）参考 GROLMES 等人（1968 年）推断的结果，提出了内部流量与 SI 工作范围之间的关系。虽然这些现有研究提出了内部流动与工作范围之间的某种关系，但水射流的稳定性与 SI 行为之间的关系尚未明确。因此，本研究旨在阐明水射流行为对 SI 工作范围的影响。我们改变了水射流出口的锥角，并研究了水射流行为与工作范围之间的关系。在这项研究中，实验使用了两种类型的水射流出口锥角：20 度和 7 度。通过对 SI 内部的详细观察，证实锥角为 7 度的水射流比锥角为 20 度的水射流的扩散性要小。在本实验的工作范围内，改变水射流的锥角没有观察到明显的差异。不过，有观点认为，具有一定分散度的水射流由于表面积大，热交换效果可能更好。这些结果表明，水射流的稳定性是运行的必要条件，但不是充分条件。
{"title":"Effect of Water Jet Behavior on the Operating Range in Supersonic Steam Injector","authors":"Yukiya Minamizono, A. Kaneko","doi":"10.1115/icone29-90414","DOIUrl":"https://doi.org/10.1115/icone29-90414","url":null,"abstract":"\u0000 In the case of a severe accident, such as a loss of power supply, it is necessary to keep the nuclear power plant in a safe condition. Therefore, steam-water injectors (SI) have been attracting attention as a device that can cool the plant without power supply in the event of a severe accident. In a SI, the steam condenses by direct contact between steam and water interface. As the steam condenses, negative pressure is generated, and a force that attracts water is generated. The SI can process a large amount of cooling water with this force. It can also function as a reactor condenser for nuclear reactors due to its high condensation performance. There are some characteristics of SI that the discharge pressure is expected to exceed the inlet pressure after being accelerated at the throat, and it has a very simple structure with converging-diverging part, which is expected to reduce the cost of installation and maintenance. Previous studies have indicated that GROLMES et al. (1968) inferred the flow in the reduced test section by measuring the void fraction and MIWA et al. (2018) suggested a relationship between internal flow and the operating range of SI, referring to the results inferred by GROLMES et al. (1968). Although these existing studies have suggested some relationship between the internal flow and the operating range, the relationship between the stability of the water jet and the behavior of the SI has not been clarified.\u0000 Therefore, the aim of this study is to clarify the effect of water jet behavior on the operating range of SI. The taper angle of the water jet outlet was varied and the relationship between the water jet behavior and the operating range was investigated. In this study, experiments were conducted using two types of taper angles at the water jet outlet: 20 and 7 degrees. As a result of detailed observation inside the SI, it was confirmed that the water jet with the taper angle of 7 degrees was less spreading than that with the taper angle of 20 degrees. In the operating range of this experiment, no significant difference was observed when the taper angle of the water jet was changed. However, it is suggested that the water jet with a certain degree of dispersion might be better in terms of heat exchange due to large surface area. These results suggest that the stability of the water jet is a necessary but not a sufficient condition for operation.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"46 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127452709","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}
Lead-cooled fast reactors are an important choice for small-scaled reactor in floating nuclear power plants. Compared with land-based nuclear power plants, floating nuclear power plants have advantages in stability, which eliminate the impact from earthquake. One aspect remains to be solved is that the lead-cooled fast reactors contains a large volume of liquid that may slosh under the marine environment. The sloshing of liquid may cause external pressure on the nuclear reactor vessel and internal components., which has a great impact on the safety of the reactor. To study the fluid structure interaction between liquid and reactor vessel, this paper establishes an numerical model for the mechanical analysis of the main vessel of a floating nuclear power plant under loads of the marine environment. The mechanical response of the main vessel of a floating nuclear power plant in the marine environment is analyzed by applying the dynamic mesh method and VOF model. The motion of marine environmental loads are applied by the remote displacement method. We study several typical loadings in operation conditions. The study shows that the stress on reactor vessel caused by fluid sloshing is relatively small when long-period loading is applied; and the effect of swaying is larger than the rolling.
{"title":"Flow-Solid Coupling Analysis of Pressure Vessels on Floating Nuclear Power Plants","authors":"Yuchao Wang, D. Lu, Fei Zhao, Yu Liu","doi":"10.1115/icone29-92039","DOIUrl":"https://doi.org/10.1115/icone29-92039","url":null,"abstract":"Lead-cooled fast reactors are an important choice for small-scaled reactor in floating nuclear power plants. Compared with land-based nuclear power plants, floating nuclear power plants have advantages in stability, which eliminate the impact from earthquake. One aspect remains to be solved is that the lead-cooled fast reactors contains a large volume of liquid that may slosh under the marine environment. The sloshing of liquid may cause external pressure on the nuclear reactor vessel and internal components., which has a great impact on the safety of the reactor. To study the fluid structure interaction between liquid and reactor vessel, this paper establishes an numerical model for the mechanical analysis of the main vessel of a floating nuclear power plant under loads of the marine environment. The mechanical response of the main vessel of a floating nuclear power plant in the marine environment is analyzed by applying the dynamic mesh method and VOF model. The motion of marine environmental loads are applied by the remote displacement method. We study several typical loadings in operation conditions. The study shows that the stress on reactor vessel caused by fluid sloshing is relatively small when long-period loading is applied; and the effect of swaying is larger than the rolling.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130446415","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}
Feiyan Dong, Shi Chen, K. Demachi, Masanori Yoshikawa, A. Seki, Shigeru Takaya
Condition monitoring is essential to the management and maintenance of Nuclear Power Plants (NPPs), as anomalies in the condition of components can affect the normal operation state of the entire plant. Therefore, timely and automatic detection of anomalies plays an important role and is in high demand. At present, deep learning is widely used for anomaly detection. Nevertheless, anomalies are difficult to define, sparsely occurring, and are accompanied by variable noise labels, which poses challenges to detection. Moreover, the problems such as loss of temporal features and gradient vanishing that exist in general deep learning models when dealing with time series data also increase the difficulty of anomaly detection. In response to these problems, a weakly supervised time series analysis framework for anomaly detection in NPPs is proposed, constituted of weakly supervised learning (WSL) and attention mechanism. The validation of the proposed framework was performed on the High Temperature Engineering Test Reactor (HTTR) anomaly cases dataset using the analytical code “ACCORD”, which distributed anomalies independently across multiple instruments and were recorded from the responding sensors to each anomaly. At this stage, 3 classes of anomalies were used as input data for the validation experiments. The experimental results demonstrate the effectiveness and feasibility of the proposed framework on anomaly tasks for the condition monitoring of NPPs.
{"title":"A Weakly Supervised Time Series Analysis Framework for Anomaly Detection in Nuclear Power Plants","authors":"Feiyan Dong, Shi Chen, K. Demachi, Masanori Yoshikawa, A. Seki, Shigeru Takaya","doi":"10.1115/icone29-91609","DOIUrl":"https://doi.org/10.1115/icone29-91609","url":null,"abstract":"\u0000 Condition monitoring is essential to the management and maintenance of Nuclear Power Plants (NPPs), as anomalies in the condition of components can affect the normal operation state of the entire plant. Therefore, timely and automatic detection of anomalies plays an important role and is in high demand. At present, deep learning is widely used for anomaly detection. Nevertheless, anomalies are difficult to define, sparsely occurring, and are accompanied by variable noise labels, which poses challenges to detection. Moreover, the problems such as loss of temporal features and gradient vanishing that exist in general deep learning models when dealing with time series data also increase the difficulty of anomaly detection. In response to these problems, a weakly supervised time series analysis framework for anomaly detection in NPPs is proposed, constituted of weakly supervised learning (WSL) and attention mechanism. The validation of the proposed framework was performed on the High Temperature Engineering Test Reactor (HTTR) anomaly cases dataset using the analytical code “ACCORD”, which distributed anomalies independently across multiple instruments and were recorded from the responding sensors to each anomaly. At this stage, 3 classes of anomalies were used as input data for the validation experiments. The experimental results demonstrate the effectiveness and feasibility of the proposed framework on anomaly tasks for the condition monitoring of NPPs.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133158063","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 heat pipe is the essential component in the heat pipe cooled reactor, it conducts heat through the phase change of the working medium and the circulating flow of gas and liquid. One end of the heat pipe is inserted into the reactor as the evaporation section, while the other end serves as the condensation section connecting to the thermoelectric conversion system. Its heat transfer capacity and temperature distribution under a steady state affect the safety, core temperature and energy conversion efficiency of the heat pipe cooled reactor system. Predicting the heat transfer process and steady-state temperature distribution of the heat pipe is helpful to learn about the heat transfer mechanism inside the heat pipe and key factors affecting its performance. It is also helpful to predict the operating characteristics of the heat pipe cooled reactor system and improve the performance of the heat pipe. The heat pipe is generally considered to be a heat transfer element with good isothermal properties. Past studies, however, proved the significant quantity of its phase-change thermal resistance, resulting in a temperature difference at the gas-liquid interface of the heat pipe that affects the evaporation and condensation. Studies on the relationship between the temperature difference, the working temperature, and heat transfer power of the heat pipe are conductive to building a heat pipe model, to predict the temperature distribution of the heat pipe in a steady state in a much more accurate manner. This paper offered a heat pipe model that applies the temperature difference of the phase change interface to calculate the evaporation and condensation. Meanwhile, a 1-meter-long sodium heat pipe was adopted to carry out steady-state heat transfer experiments. Measured and computed the temperature difference of the phase change interface of the heat pipe under varying power and temperature and conducted the relational equation as the input of the heat pipe model. The model predicted the steady-state temperature distribution of the sodium heat pipe and the transient when the power changes after the steam reach the continuous flow stage and compared the calculated outcomes with the experimental values. The transient error was less than 20K, and the steady-state error was less than 12K. The results show that the temperature difference at the phase change interface has a great influence on the steady-state temperature distribution of the heat pipe, which changes with the operating temperature. However, the physical and geometric factors affecting the temperature difference need to be further studied, to reduce the temperature difference as much as possible and improve the energy conversion efficiency of the heat pipe reactor system in the future.
{"title":"A Transient Model of Sodium Heat Pipe With Phase Change Temperature Difference","authors":"Ruicheng Zhong, Yugao Ma, Qingzhu Zhao, Jian Deng, Yu Liu, Shuhua Ding","doi":"10.1115/icone29-92202","DOIUrl":"https://doi.org/10.1115/icone29-92202","url":null,"abstract":"\u0000 The heat pipe is the essential component in the heat pipe cooled reactor, it conducts heat through the phase change of the working medium and the circulating flow of gas and liquid. One end of the heat pipe is inserted into the reactor as the evaporation section, while the other end serves as the condensation section connecting to the thermoelectric conversion system. Its heat transfer capacity and temperature distribution under a steady state affect the safety, core temperature and energy conversion efficiency of the heat pipe cooled reactor system. Predicting the heat transfer process and steady-state temperature distribution of the heat pipe is helpful to learn about the heat transfer mechanism inside the heat pipe and key factors affecting its performance. It is also helpful to predict the operating characteristics of the heat pipe cooled reactor system and improve the performance of the heat pipe. The heat pipe is generally considered to be a heat transfer element with good isothermal properties. Past studies, however, proved the significant quantity of its phase-change thermal resistance, resulting in a temperature difference at the gas-liquid interface of the heat pipe that affects the evaporation and condensation. Studies on the relationship between the temperature difference, the working temperature, and heat transfer power of the heat pipe are conductive to building a heat pipe model, to predict the temperature distribution of the heat pipe in a steady state in a much more accurate manner. This paper offered a heat pipe model that applies the temperature difference of the phase change interface to calculate the evaporation and condensation. Meanwhile, a 1-meter-long sodium heat pipe was adopted to carry out steady-state heat transfer experiments. Measured and computed the temperature difference of the phase change interface of the heat pipe under varying power and temperature and conducted the relational equation as the input of the heat pipe model. The model predicted the steady-state temperature distribution of the sodium heat pipe and the transient when the power changes after the steam reach the continuous flow stage and compared the calculated outcomes with the experimental values. The transient error was less than 20K, and the steady-state error was less than 12K. The results show that the temperature difference at the phase change interface has a great influence on the steady-state temperature distribution of the heat pipe, which changes with the operating temperature. However, the physical and geometric factors affecting the temperature difference need to be further studied, to reduce the temperature difference as much as possible and improve the energy conversion efficiency of the heat pipe reactor system in the future.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133712724","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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