Fault Tree Analysis (FTA) is a powerful and well-established tool, widely-used to evaluate system reliability. The logical connections between faults and causes in Fault Trees (FT) are assumed to be deterministic and are represented graphically via logical gates (such as AND gate, OR gate, NOT gate, etc.). However, sometimes the causalities can be uncertain. Considering that some of the causal relationships in FTs may be uncertain or non-deterministic, we propose a new model to represent the uncertainties, so called as Condition Fault Tree (CFT). We extend the traditional FTA by introducing a new parameter U, which illustrates the random mechanism of how parent event can cause child event. The probability of U (which is denoted by u = Pr{U}), is used to measure the uncertainty between parent event and child event. By introducing rules of parameter U in CFT, we explore its properties and corollaries. We also introduce a methodology to simplify CFTs based on Contraction, Elimination and Extraction rules. With the simplification rules, the structure of CFT can be simplified and the size of CFT can be significantly reduced. Since CFT is an extension of traditional FT, a qualitative analysis method and a quantitative method are introduced. For qualitative analysis, one can simplify a given CFT into the simplest form with the aforementioned rules, properties, and corollaries. With the simplest form of CFT, one can then get the Minimum Cut Sets with uncertainties, as an extension of Minimum Cut Sets. For quantitative analysis, exact calculation methods based on Inclusion-Exclusion and Disjoint-Sum-of-Product are proposed. Some examples are used to illustrate how CFT works.
{"title":"Condition Fault Tree: An Extension of Traditional Fault Tree to Handle Uncertainty","authors":"Zhenxu Zhou, Qin Zhang","doi":"10.1115/ICONE26-81243","DOIUrl":"https://doi.org/10.1115/ICONE26-81243","url":null,"abstract":"Fault Tree Analysis (FTA) is a powerful and well-established tool, widely-used to evaluate system reliability. The logical connections between faults and causes in Fault Trees (FT) are assumed to be deterministic and are represented graphically via logical gates (such as AND gate, OR gate, NOT gate, etc.). However, sometimes the causalities can be uncertain. Considering that some of the causal relationships in FTs may be uncertain or non-deterministic, we propose a new model to represent the uncertainties, so called as Condition Fault Tree (CFT). We extend the traditional FTA by introducing a new parameter U, which illustrates the random mechanism of how parent event can cause child event. The probability of U (which is denoted by u = Pr{U}), is used to measure the uncertainty between parent event and child event. By introducing rules of parameter U in CFT, we explore its properties and corollaries. We also introduce a methodology to simplify CFTs based on Contraction, Elimination and Extraction rules. With the simplification rules, the structure of CFT can be simplified and the size of CFT can be significantly reduced. Since CFT is an extension of traditional FT, a qualitative analysis method and a quantitative method are introduced. For qualitative analysis, one can simplify a given CFT into the simplest form with the aforementioned rules, properties, and corollaries. With the simplest form of CFT, one can then get the Minimum Cut Sets with uncertainties, as an extension of Minimum Cut Sets. For quantitative analysis, exact calculation methods based on Inclusion-Exclusion and Disjoint-Sum-of-Product are proposed. Some examples are used to illustrate how CFT works.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"7 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":"133412263","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}
Knowledge of the heat transfer phenomenon under flow decay transient condition is important for the safety assessment of a very high temperature reactor (VHTR) during a loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally investigated. The experiment was conducted by using a forced convection heat transfer experimental apparatus. A flow control value with its control system was used to realize a flow decay condition. Helium gas was used as a coolant, and a platinum cylinder with a diameter of 1 mm was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Subsequently, the flow rate of the helium gas began to exponentially decrease with different time constants ranging from 3 s to 15 s. The initial flow velocity ranged from 7 m/s to 10 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under a wide range of experimental conditions such as heat generation rates and flow decay time constants. The results indicated that the temperature of the test heater exhibits a rapid increase during this process, and the increasing rate of the temperature is higher for a lower time constant. An increase in the heat generation rate leads to a higher increase in the surface temperature. Therefore, the heat generation rates of the fuel rods are high when a VHTR operates at high power, and it is more challenging to implement passive safety design to ensure the temperature limitation of the fuel rods during a loss-of-coolant accident. Moreover, the heat transfer coefficient relative to time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was examined based on the experimental data.
{"title":"Transient Heat Transfer for Helium Gas at Various Flow Decay Time Constants and Heat Generation Rates","authors":"Qiusheng Liu, A. Kitano, K. Fukuda, M. Shibahara","doi":"10.1115/ICONE26-81391","DOIUrl":"https://doi.org/10.1115/ICONE26-81391","url":null,"abstract":"Knowledge of the heat transfer phenomenon under flow decay transient condition is important for the safety assessment of a very high temperature reactor (VHTR) during a loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally investigated. The experiment was conducted by using a forced convection heat transfer experimental apparatus. A flow control value with its control system was used to realize a flow decay condition. Helium gas was used as a coolant, and a platinum cylinder with a diameter of 1 mm was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Subsequently, the flow rate of the helium gas began to exponentially decrease with different time constants ranging from 3 s to 15 s. The initial flow velocity ranged from 7 m/s to 10 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under a wide range of experimental conditions such as heat generation rates and flow decay time constants. The results indicated that the temperature of the test heater exhibits a rapid increase during this process, and the increasing rate of the temperature is higher for a lower time constant. An increase in the heat generation rate leads to a higher increase in the surface temperature. Therefore, the heat generation rates of the fuel rods are high when a VHTR operates at high power, and it is more challenging to implement passive safety design to ensure the temperature limitation of the fuel rods during a loss-of-coolant accident. Moreover, the heat transfer coefficient relative to time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was examined based on the experimental data.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"7 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":"133174781","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 phenomenon that bubble bursts at the water surface and results in droplets production is one of the source of radioactive aerosol release, when the gas goes through the aerosol pool. Based on this, a high-speed photographic visualization experimental device was used to visualize the bubble bursting process at liquid surface under different conditions. Experimental studies show that: the bursting process of the bubbles with 7mm–28mm in diameter is a Single point rupture process. The process includes bubble rising, bubble cap draining, punctured point appearing, the liquid film rolling-up which forms the liquid ring, droplets emission as liquid ring breaks. The different punctured position changes the process of bubble bursting and the distribution of the droplets, thus the location of punctured position were divided into different area, which mainly locates at the foot of the bubble cap. Furthermore, the change of liquid phase conditions will affects the location of the punctured position, the number and the sizes of droplets. In the experiments, as temperature of the liquid phase changes from 16°C to 60°C, the process of drainage of bubble cap is shortened, and the probability of punctured position at the bottom increases. When punctured position is the same position, the number of droplets decreased and the diameter of droplet increased as temperature was increasing.
{"title":"Experimental Study on Bubble Bursting and Droplet Releasing Characteristics Under Different Liquid Phase Conditions","authors":"Hao Chen, Haifeng Gu, Xiang Yu, Yanmin Zhou, Zhong-ning Sun, Jiming Wen","doi":"10.1115/ICONE26-82047","DOIUrl":"https://doi.org/10.1115/ICONE26-82047","url":null,"abstract":"The phenomenon that bubble bursts at the water surface and results in droplets production is one of the source of radioactive aerosol release, when the gas goes through the aerosol pool. Based on this, a high-speed photographic visualization experimental device was used to visualize the bubble bursting process at liquid surface under different conditions. Experimental studies show that: the bursting process of the bubbles with 7mm–28mm in diameter is a Single point rupture process. The process includes bubble rising, bubble cap draining, punctured point appearing, the liquid film rolling-up which forms the liquid ring, droplets emission as liquid ring breaks. The different punctured position changes the process of bubble bursting and the distribution of the droplets, thus the location of punctured position were divided into different area, which mainly locates at the foot of the bubble cap. Furthermore, the change of liquid phase conditions will affects the location of the punctured position, the number and the sizes of droplets. In the experiments, as temperature of the liquid phase changes from 16°C to 60°C, the process of drainage of bubble cap is shortened, and the probability of punctured position at the bottom increases. When punctured position is the same position, the number of droplets decreased and the diameter of droplet increased as temperature was increasing.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"38 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":"129214721","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}
Naoto Kitahara, Yasunori Yamamoto, T. Narabayashi, G. Chiba
Two-phase flow experiments and analysis were conducted to understand two-phase flow behavior of the water scrubbing pool of the filtered containment venting system with steam injection. In the early phase of steam injection, the water level gradually increased due to the steam condensation where the water surface was stable. After the water pool reached the saturation temperature, the diameter of bubbles increased when the bubbles moved upward in the water pool, where fluctuation of the water surface was observed. The water level increased when the scrubbing pool was pressurized by an orifice. Our simulation results showed that the decrement of the bubble velocity due to the pressurization may promoted the level swell.
{"title":"Pressure Dependence of Two Phase Flow Behavior of Stagnant Water in a Vertical Pipe During Steam Injection","authors":"Naoto Kitahara, Yasunori Yamamoto, T. Narabayashi, G. Chiba","doi":"10.1115/ICONE26-82545","DOIUrl":"https://doi.org/10.1115/ICONE26-82545","url":null,"abstract":"Two-phase flow experiments and analysis were conducted to understand two-phase flow behavior of the water scrubbing pool of the filtered containment venting system with steam injection. In the early phase of steam injection, the water level gradually increased due to the steam condensation where the water surface was stable. After the water pool reached the saturation temperature, the diameter of bubbles increased when the bubbles moved upward in the water pool, where fluctuation of the water surface was observed. The water level increased when the scrubbing pool was pressurized by an orifice. Our simulation results showed that the decrement of the bubble velocity due to the pressurization may promoted the level swell.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"67 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":"115408541","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. Aly, V. Petrov, M. Avramova, A. Manera, K. Ivanov
The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Eventually, its concentration can increase beyond a certain limit above which hydrogen precipitates as hydrides. The formation of hydrides can embrittle the cladding and leads to micro-cracks that can compromise the cladding integrity. At the spacer grids locations, the mixing vanes will create swirl flow and mixing of the coolant leading to a high temperature gradient on the fuel rod cladding. This temperature gradient is a strong driving force for hydrogen to diffuse from high to low temperature locations. Therefore, the hydrogen behavior around the spacer grids with mixing vanes is important to model. In this work, the computational fluid dynamics code START-CCM+ is used to model the effect of the mixing vanes on the temperature profile on the cladding outer surface. It ws coupled with the transport code MPACT and the fuel performance code BISON. The computational model consisted of a 5 × 5 fuel rods subassembly with a guide tube in the central location. The obtained cladding temperature profile on a fuel rod of interest was applied as a boundary condition to BISON to model the hydrogen behavior around the spacer grids in a three-dimensional manner. Three spacer grids were modeled at elevations of 217.9 cm, 270.14 cm and 322.35 cm. The hydrogen behavior at each of those locations is evaluated and compared to assess the importance order of those locations.
{"title":"Evaluation of the Mixing Vanes Effect on the Hydrogen Diffusion and Hydride Formation in the Fuel Cladding","authors":"A. Aly, V. Petrov, M. Avramova, A. Manera, K. Ivanov","doi":"10.1115/ICONE26-82431","DOIUrl":"https://doi.org/10.1115/ICONE26-82431","url":null,"abstract":"The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Eventually, its concentration can increase beyond a certain limit above which hydrogen precipitates as hydrides. The formation of hydrides can embrittle the cladding and leads to micro-cracks that can compromise the cladding integrity.\u0000 At the spacer grids locations, the mixing vanes will create swirl flow and mixing of the coolant leading to a high temperature gradient on the fuel rod cladding. This temperature gradient is a strong driving force for hydrogen to diffuse from high to low temperature locations. Therefore, the hydrogen behavior around the spacer grids with mixing vanes is important to model. In this work, the computational fluid dynamics code START-CCM+ is used to model the effect of the mixing vanes on the temperature profile on the cladding outer surface. It ws coupled with the transport code MPACT and the fuel performance code BISON. The computational model consisted of a 5 × 5 fuel rods subassembly with a guide tube in the central location. The obtained cladding temperature profile on a fuel rod of interest was applied as a boundary condition to BISON to model the hydrogen behavior around the spacer grids in a three-dimensional manner. Three spacer grids were modeled at elevations of 217.9 cm, 270.14 cm and 322.35 cm. The hydrogen behavior at each of those locations is evaluated and compared to assess the importance order of those locations.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"83 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":"115610088","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}
Back-to-back pipe bends are widely adopted applications in many industries including nuclear sectors. Evaluation of their load bearing capability under complex cyclic loading is very important. Recently, a couple of research reported shakedown boundary of a 90° back-to-back pipe bends by adopting a conservative approach but no comprehensive post yield structural behaviors have been dealt with. In this research the concerning pipe bends subjected to cyclic opening in-plane (IP)/out-of-plane (OP) bending and steady internal pressures are analyzed to construct shakedown and ratchet limit boundary by means of the Linear Matching Method. Analyzed results present that the concerning pipe bends under out-of-plane bending has higher resistance to cyclic bending than under in-plane bending. In additions, the out-of-plane bending causes very small alternating plasticity areas, unlike the in-plane bending. Full cyclic incremental analyses known as step-by-step analysis are performed to verify the structural responses either side of each boundary and confirm correct responses. Parametric studies are carried out with respect to changes in geometry of the concerning pipe bends subjected to the same loading, and semi-empirical equations are derived from relationships of the reverse plasticity limit and the limit pressure with the bend characteristic. This paper offers broad understandings of structural responses of the 90° back-to-back pipe bends under the complex cyclic loading as well as providing key points to be considered for the life assessment of the piping system.
{"title":"Cyclic Plasticity Behavior of 90° Back-to-Back Pipe Bends Under Cyclic Bending and Steady Pressure","authors":"N. Cho, Haofeng Chen","doi":"10.1115/ICONE26-82386","DOIUrl":"https://doi.org/10.1115/ICONE26-82386","url":null,"abstract":"Back-to-back pipe bends are widely adopted applications in many industries including nuclear sectors. Evaluation of their load bearing capability under complex cyclic loading is very important. Recently, a couple of research reported shakedown boundary of a 90° back-to-back pipe bends by adopting a conservative approach but no comprehensive post yield structural behaviors have been dealt with. In this research the concerning pipe bends subjected to cyclic opening in-plane (IP)/out-of-plane (OP) bending and steady internal pressures are analyzed to construct shakedown and ratchet limit boundary by means of the Linear Matching Method. Analyzed results present that the concerning pipe bends under out-of-plane bending has higher resistance to cyclic bending than under in-plane bending. In additions, the out-of-plane bending causes very small alternating plasticity areas, unlike the in-plane bending. Full cyclic incremental analyses known as step-by-step analysis are performed to verify the structural responses either side of each boundary and confirm correct responses. Parametric studies are carried out with respect to changes in geometry of the concerning pipe bends subjected to the same loading, and semi-empirical equations are derived from relationships of the reverse plasticity limit and the limit pressure with the bend characteristic. This paper offers broad understandings of structural responses of the 90° back-to-back pipe bends under the complex cyclic loading as well as providing key points to be considered for the life assessment of the piping system.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"662 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":"122964985","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 sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety. The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi‘an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.
{"title":"Numerical and Experimental Investigation on Core Assembly Thermal-Gradient-Induced Deformation of Sodium-Cooled Fast Reactor","authors":"Ma Zehua, Yingwei Wu, G. Su, W. Tian, S. Qiu","doi":"10.1115/ICONE26-81891","DOIUrl":"https://doi.org/10.1115/ICONE26-81891","url":null,"abstract":"In sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety. The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi‘an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"5 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":"126150495","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}
Yandong Hou, Liu Wang, Yingwei Wu, W. Tian, S. Qiu, G. Su
Pressure drop experiments was conducted for liquid sodium in an electrically heated 7-rod bundle. The electrically heated 7-rod bundle was placed in a hexagonal tube. In the experiment, the heat flux ranges from 0∼300 kw · m−2, mass velocity from 40∼450 kg · m−2 · s−1, system pressure from 10∼200 KPa and the average temperature of liquid sodium from 350∼650°C. The effects of the heat flux, system pressure and the average temperature of liquid sodium on the pressure drop was in-depth analyzed. A new correlation for pressure drop was developed based on the experimental data of liquid sodium in a 7-rod bundle.
{"title":"Pressure Drop Experiments of Liquid Sodium Flowing in a 7-Rod Bundle","authors":"Yandong Hou, Liu Wang, Yingwei Wu, W. Tian, S. Qiu, G. Su","doi":"10.1115/ICONE26-81444","DOIUrl":"https://doi.org/10.1115/ICONE26-81444","url":null,"abstract":"Pressure drop experiments was conducted for liquid sodium in an electrically heated 7-rod bundle. The electrically heated 7-rod bundle was placed in a hexagonal tube. In the experiment, the heat flux ranges from 0∼300 kw · m−2, mass velocity from 40∼450 kg · m−2 · s−1, system pressure from 10∼200 KPa and the average temperature of liquid sodium from 350∼650°C. The effects of the heat flux, system pressure and the average temperature of liquid sodium on the pressure drop was in-depth analyzed. A new correlation for pressure drop was developed based on the experimental data of liquid sodium in a 7-rod bundle.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"6 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":"121692155","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}
Steam generator is a key component for realizing the heat exchange between the secondary sodium circuit and the tertiary circuit water in the sodium-cooled fast reactor. Any tiny crack of the heat-transfer tube in the steam generator may lead to the contact of water and sodium, resulting in the sodium-water reaction and the generation of hydrogen, reaction heat and other corrosive products, which further corrode the break and the adjacent tubes. In order to effectively curb the further development of small leak, timely detection of the occurrence of sodium-water reactions is particularly important. With the background of the small leak sodium water reaction caused by the crack of the steam generator heat transfer tube, in this paper the migration behavior of hydrogen in the secondary sodium circuit is analyzed, based on the mass, momentum and energy conservation equations, and the concentration of hydrogen in the secondary sodium circuit is numerically calculated one-dimensionally. The results show that the hydrogen concentration at a certain point in the circuit increases with the increase of water leaking time as a whole, when the water injection is intermittent, the hydrogen concentration at a certain point in the circuit fluctuates periodically and the basal concentration becomes higher and higher, which provides a reference for the arrangement of the secondary loop small leak detection system.
{"title":"Study on Hydrogen Migration in Small Water Leak of Sodium-Cooled Fast Reactor","authors":"Xinjie Deng, Xuewu Cao","doi":"10.1115/ICONE26-81314","DOIUrl":"https://doi.org/10.1115/ICONE26-81314","url":null,"abstract":"Steam generator is a key component for realizing the heat exchange between the secondary sodium circuit and the tertiary circuit water in the sodium-cooled fast reactor. Any tiny crack of the heat-transfer tube in the steam generator may lead to the contact of water and sodium, resulting in the sodium-water reaction and the generation of hydrogen, reaction heat and other corrosive products, which further corrode the break and the adjacent tubes. In order to effectively curb the further development of small leak, timely detection of the occurrence of sodium-water reactions is particularly important. With the background of the small leak sodium water reaction caused by the crack of the steam generator heat transfer tube, in this paper the migration behavior of hydrogen in the secondary sodium circuit is analyzed, based on the mass, momentum and energy conservation equations, and the concentration of hydrogen in the secondary sodium circuit is numerically calculated one-dimensionally. The results show that the hydrogen concentration at a certain point in the circuit increases with the increase of water leaking time as a whole, when the water injection is intermittent, the hydrogen concentration at a certain point in the circuit fluctuates periodically and the basal concentration becomes higher and higher, which provides a reference for the arrangement of the secondary loop small leak detection system.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"14 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":"124027685","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}
Currently, nuclear power plays a quite visible role in the world electricity generation (∼11%). However, before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011, NPPs generated about 14% of the world’s electricity. Accounting that after, mainly, Chernobyl NPP severe accident a number of power reactors built and put into operation in the world decreased from 120 within 1985–1990 to about 22 per 5 years (within 1995–2015), we might face a significant shortage of operating power reactors within 2030–2040. Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 years) and far away (10–25 years and beyond) future trends in nuclear-power industry. In the current paper statistics on all current nuclear-power reactors were analyzed and based on that future trends were estimated in terms of types of reactors to be left after 10 years, new types of reactors to be put into operation, projections of how many reactors and of which types will be build. To make any projections an average operating term of power reactors should be estimated. In the current paper a nuclear-power-reactor operating term of 45 years was considered. Also, rates of building and putting into operation power reactors worldwide were estimated, and several scenarios of future developments in nuclear-power industry in the world and in selected countries were considered.
{"title":"Study on Current Status and Future Developments in Nuclear-Power Industry of the World","authors":"R. Pioro, A. Zvorykin, R. Machrafi, I. Pioro","doi":"10.1115/ICONE26-82085","DOIUrl":"https://doi.org/10.1115/ICONE26-82085","url":null,"abstract":"Currently, nuclear power plays a quite visible role in the world electricity generation (∼11%). However, before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011, NPPs generated about 14% of the world’s electricity. Accounting that after, mainly, Chernobyl NPP severe accident a number of power reactors built and put into operation in the world decreased from 120 within 1985–1990 to about 22 per 5 years (within 1995–2015), we might face a significant shortage of operating power reactors within 2030–2040.\u0000 Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 years) and far away (10–25 years and beyond) future trends in nuclear-power industry. In the current paper statistics on all current nuclear-power reactors were analyzed and based on that future trends were estimated in terms of types of reactors to be left after 10 years, new types of reactors to be put into operation, projections of how many reactors and of which types will be build.\u0000 To make any projections an average operating term of power reactors should be estimated. In the current paper a nuclear-power-reactor operating term of 45 years was considered. Also, rates of building and putting into operation power reactors worldwide were estimated, and several scenarios of future developments in nuclear-power industry in the world and in selected countries were considered.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"64 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":"127617795","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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