Nakamura Yuki, Kota Fujiwara, Wataru Kikuchi, Shimpei Saito, T. Yuasa, A. Kaneko, Y. Abe
In severe accidents of nuclear power plants, large amounts of gas containing radioactive particles are generated. In the process of gas release into the atmosphere, it is needed to suppress the leakage of radioactive material. The gas is decontaminated by moving radioactive particles from the gas phase to the liquid phase. This effect of capturing particles is called pool scrubbing, and it has been verified great decontamination effect. Therefore, it is extremely important to analyze the effect in evaluating the influence to the environment. But study on its principle is not carried out sufficiently. And also we don’t have enough experimental date to analyze the effect. The purpose of this study is to clarify the gas-liquid two-phase flow behavior which is important in elucidating the mechanism of pool scrubbing. Particularly, this study focused on the behavior of bubble generation and breakup after being injected from the nozzle and the flow structure of rising bubbles in the still water. Furthermore, we evaluate the validity of the model used in the existing severe accidents analysis code such as the MELCOR by comparing the model with experimental data. We measure the gas phase jet injected from the upward nozzle inserted to a test water tank. Nozzle diameter, gas phase flow rate, liquid phase temperature, and water depth were used as parameters. Bubble behavior was observed via a high-speed camera. The bubble rising speed, bubble distribution and void fraction were measured by a wire mesh sensor. In previous studies, experiments using non-condensable gas in normal temperature water have been mainly conducted. In order to conduct the experiment under conditions that simulate actual equipment, steam which is a condensable gas was used in this study. Moreover, it is assumed that thermal stratification is formed in the pressure suppression pool during severe accidents. To reproduce this situation, thermal stratification is formed in the test water tank. For bubble behavior and flow phenomenon, the result of using non-condensable gas was compared with that using steam. We consider the influence of formation of a thermal stratification. As described above, the flow phenomenon in the pool scrubbing was visualized and measured. Finally, we discuss the validity of the analysis code by comparing the calculation formula and model in the analysis code with the experiment data.
{"title":"Two Phase Flow Behavior During Pool Scrubbing","authors":"Nakamura Yuki, Kota Fujiwara, Wataru Kikuchi, Shimpei Saito, T. Yuasa, A. Kaneko, Y. Abe","doi":"10.1115/ICONE26-81497","DOIUrl":"https://doi.org/10.1115/ICONE26-81497","url":null,"abstract":"In severe accidents of nuclear power plants, large amounts of gas containing radioactive particles are generated. In the process of gas release into the atmosphere, it is needed to suppress the leakage of radioactive material. The gas is decontaminated by moving radioactive particles from the gas phase to the liquid phase. This effect of capturing particles is called pool scrubbing, and it has been verified great decontamination effect. Therefore, it is extremely important to analyze the effect in evaluating the influence to the environment. But study on its principle is not carried out sufficiently. And also we don’t have enough experimental date to analyze the effect.\u0000 The purpose of this study is to clarify the gas-liquid two-phase flow behavior which is important in elucidating the mechanism of pool scrubbing. Particularly, this study focused on the behavior of bubble generation and breakup after being injected from the nozzle and the flow structure of rising bubbles in the still water. Furthermore, we evaluate the validity of the model used in the existing severe accidents analysis code such as the MELCOR by comparing the model with experimental data.\u0000 We measure the gas phase jet injected from the upward nozzle inserted to a test water tank. Nozzle diameter, gas phase flow rate, liquid phase temperature, and water depth were used as parameters. Bubble behavior was observed via a high-speed camera. The bubble rising speed, bubble distribution and void fraction were measured by a wire mesh sensor. In previous studies, experiments using non-condensable gas in normal temperature water have been mainly conducted. In order to conduct the experiment under conditions that simulate actual equipment, steam which is a condensable gas was used in this study. Moreover, it is assumed that thermal stratification is formed in the pressure suppression pool during severe accidents. To reproduce this situation, thermal stratification is formed in the test water tank. For bubble behavior and flow phenomenon, the result of using non-condensable gas was compared with that using steam. We consider the influence of formation of a thermal stratification. As described above, the flow phenomenon in the pool scrubbing was visualized and measured. Finally, we discuss the validity of the analysis code by comparing the calculation formula and model in the analysis code with the experiment data.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"11 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":"126887101","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}
This study aims to investigate the coupling dependence resulting from three and four operating parameters for burnup credit calculations in boiling water reactor (BWR) spent fuel assemblies. Four operating parameters are under investigation, including fuel temperature, axial burnup profile, axial moderator density profile and control blade usage. In this study, the effects of variation on the curve of effective multiplication factor (keff) versus burnup (B) resulting from one and multiple operating parameters were defined as the single and compound effects, respectively. Particularly, the compound effects adopt more practical operating parameters than single effects does and thus affect the precise assessment to some extent. In our previous study, the compound effects resulting from two operating parameters were investigated in depth. However, the influence of compound effects resulting from three and four operating parameters on burnup credit calculation is still unknown. Therefore, this constitutes the purpose of this study. All the calculations were performed using SCALE 6.1 computer code together with the ENDF/B-VII 238 energy group neutron data library. Two geometrical models were established to represent the typical GE14 10 × 10 BWR fuel assembly and the GBC-68 storage cask. The results revealed that the reactivity deviation (or changes of keff, Δk) resulting from the compound effects was not a summation of the Δk’s resulting from the associated single effects. Moreover, such Δk discrepancies increase as B increases. In this study, the curves of keff versus B due to single and compound effects were approximated by a second degree polynomial of B. A general formula was thus proposed to express these curves.
{"title":"Coupling Dependence of Multiple Operating Parameters on Burnup Credit Calculations for BWR Spent Fuel Assemblies","authors":"Shang-Chien Wu, D. Chao, Jenq-Horng Liang","doi":"10.1115/ICONE26-82156","DOIUrl":"https://doi.org/10.1115/ICONE26-82156","url":null,"abstract":"This study aims to investigate the coupling dependence resulting from three and four operating parameters for burnup credit calculations in boiling water reactor (BWR) spent fuel assemblies. Four operating parameters are under investigation, including fuel temperature, axial burnup profile, axial moderator density profile and control blade usage. In this study, the effects of variation on the curve of effective multiplication factor (keff) versus burnup (B) resulting from one and multiple operating parameters were defined as the single and compound effects, respectively. Particularly, the compound effects adopt more practical operating parameters than single effects does and thus affect the precise assessment to some extent. In our previous study, the compound effects resulting from two operating parameters were investigated in depth. However, the influence of compound effects resulting from three and four operating parameters on burnup credit calculation is still unknown. Therefore, this constitutes the purpose of this study. All the calculations were performed using SCALE 6.1 computer code together with the ENDF/B-VII 238 energy group neutron data library. Two geometrical models were established to represent the typical GE14 10 × 10 BWR fuel assembly and the GBC-68 storage cask. The results revealed that the reactivity deviation (or changes of keff, Δk) resulting from the compound effects was not a summation of the Δk’s resulting from the associated single effects. Moreover, such Δk discrepancies increase as B increases. In this study, the curves of keff versus B due to single and compound effects were approximated by a second degree polynomial of B. A general formula was thus proposed to express these curves.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"9 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125004190","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}
This Paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for CFD analysis of VVER-1000 reactor baffle cooling is presented. Firstly, an analytical model was developed to obtain boundary conditions and simplify CFD analysis, i.e. the real geometry of the cooling channels was replaced by equivalent elements, the core was presented as porous body with the appropriate characteristics. Secondly, the CFD analysis was performed using 60–degree symmetry, which included: core, baffle and core barrel, it is limited by the height of the baffle. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes a simplified geometry of connecting studs, considering cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5/mod3.2. Obtained results were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased.
{"title":"Reactor Baffle Cooling CFD Framework for Swelling Assessment","authors":"Y. Filonova, V. Filonov, Y. Dubyk","doi":"10.1115/ICONE26-82365","DOIUrl":"https://doi.org/10.1115/ICONE26-82365","url":null,"abstract":"This Paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for CFD analysis of VVER-1000 reactor baffle cooling is presented. Firstly, an analytical model was developed to obtain boundary conditions and simplify CFD analysis, i.e. the real geometry of the cooling channels was replaced by equivalent elements, the core was presented as porous body with the appropriate characteristics. Secondly, the CFD analysis was performed using 60–degree symmetry, which included: core, baffle and core barrel, it is limited by the height of the baffle. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes a simplified geometry of connecting studs, considering cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5/mod3.2. Obtained results were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124376389","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 Transient Reactor Test Facility (TREAT) is a high enriched, graphite moderated, air cooled reactor built for experimental transient fuel testing. Recently, the reactor was returned to operation after having been shut down since 1994. Transients at TREAT are controlled largely by control/transient rod movement and temperature feedback that is attributed to the core’s graphite-fuel matrix. To date, TREAT simulations use the standard ENDF/B-VII.1 graphite thermal neutron scattering cross sections that assume an ideal crystalline form for the core’s graphite. Historically, it has been reported that the use of these cross sections may result in a −2000 pcm discrepancy when attempting to predict TREAT criticality [1]. In this work, a multi-physics simulation of a TREAT transient is performed using the standard ENDF/B-VII.1 graphite thermal scattering cross section libraries and compared with results using graphite libraries that assume a porous graphite structure and a corresponding density consistent with TREAT graphite. The transient simulation methodology couples a full-core transient Monte Carlo calculation in the Serpent code with feedback calculated from temperature estimates derived using the computational fluid dynamics code OpenFOAM. Steady state simulations show that use of the “porous” graphite libraries allows predicting TREAT criticality to within a few hundred pcm. In the current transient simulations, the reactor’s time dependent power behavior is successfully reproduced. With this model, observables such as maximum fuel temperatures and temperature-dependent flux spectra are calculated, using both the traditional ENDF/B-VII.1 and the “porous” graphite thermal scattering libraries.
{"title":"TREAT Transient Modeling and Impact of Graphite Thermal Scattering","authors":"Nina C. Sorrell, A. Hawari","doi":"10.1115/ICONE26-81887","DOIUrl":"https://doi.org/10.1115/ICONE26-81887","url":null,"abstract":"The Transient Reactor Test Facility (TREAT) is a high enriched, graphite moderated, air cooled reactor built for experimental transient fuel testing. Recently, the reactor was returned to operation after having been shut down since 1994. Transients at TREAT are controlled largely by control/transient rod movement and temperature feedback that is attributed to the core’s graphite-fuel matrix. To date, TREAT simulations use the standard ENDF/B-VII.1 graphite thermal neutron scattering cross sections that assume an ideal crystalline form for the core’s graphite. Historically, it has been reported that the use of these cross sections may result in a −2000 pcm discrepancy when attempting to predict TREAT criticality [1]. In this work, a multi-physics simulation of a TREAT transient is performed using the standard ENDF/B-VII.1 graphite thermal scattering cross section libraries and compared with results using graphite libraries that assume a porous graphite structure and a corresponding density consistent with TREAT graphite. The transient simulation methodology couples a full-core transient Monte Carlo calculation in the Serpent code with feedback calculated from temperature estimates derived using the computational fluid dynamics code OpenFOAM. Steady state simulations show that use of the “porous” graphite libraries allows predicting TREAT criticality to within a few hundred pcm. In the current transient simulations, the reactor’s time dependent power behavior is successfully reproduced. With this model, observables such as maximum fuel temperatures and temperature-dependent flux spectra are calculated, using both the traditional ENDF/B-VII.1 and the “porous” graphite thermal scattering libraries.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"35 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":"129135127","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. Zvorykin, M. Mahdi, Roman Popov, K. Barati Far, I. Pioro
Current Nuclear Power Plants (NPPs) equipped with water-cooled reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants. Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV nuclear-reactor concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs. Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required. Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.
{"title":"Heat Transfer to Supercritical Water (Liquid-Like State) Flowing in a Short Vertical Bare Tube With Upward Flow","authors":"A. Zvorykin, M. Mahdi, Roman Popov, K. Barati Far, I. Pioro","doi":"10.1115/ICONE26-81608","DOIUrl":"https://doi.org/10.1115/ICONE26-81608","url":null,"abstract":"Current Nuclear Power Plants (NPPs) equipped with water-cooled reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants.\u0000 Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV nuclear-reactor concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs.\u0000 In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs.\u0000 Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required.\u0000 Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"11 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":"126420488","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}
Secondary loop system is an essential part in nuclear power plant (NPP). The traditional methods for the secondary loop cannot reveal the universality for the most of the NPPs. When there is the need to build another NPP secondary loop simulation program or project, the simulation model always need to modify or adapt to the new design parameters and operation parameters. In this work a fast modeling simulation platform for secondary loop in NPP is developed. In this simulation platform there is the modular, parametric simulation model in the device library and graphical human-computer interaction. There is the modeling procedure for the NPP secondary loop system simulating and modeling. In the device library there are different simulation models of secondary loop system. The secondary loop simulation model can be made up by the device simulation model with reasonable data interface. Each device simulation model could represent the practical device steady and dynamic performance with accurate calculation. The simulation results calculated in the simulation platform can represent the accuracy steady calculation and reasonable dynamic tendency of main parameters for the secondary loop system built. From the simulation platform the NPP secondary loop system simulation model can be built conveniently and fast. In the simulation platform there is the appropriate data input form and output form. The simulation platform can be used in the different purposes of simulation for NPP secondary loop as training, evaluating, operation and characteristic analysis.
{"title":"Research of Fast Modeling and Simulating Platform for Nuclear Power Plant Secondary Loop","authors":"Mei Gong, M. Peng, Haishan Zhu","doi":"10.1115/ICONE26-81779","DOIUrl":"https://doi.org/10.1115/ICONE26-81779","url":null,"abstract":"Secondary loop system is an essential part in nuclear power plant (NPP). The traditional methods for the secondary loop cannot reveal the universality for the most of the NPPs. When there is the need to build another NPP secondary loop simulation program or project, the simulation model always need to modify or adapt to the new design parameters and operation parameters. In this work a fast modeling simulation platform for secondary loop in NPP is developed. In this simulation platform there is the modular, parametric simulation model in the device library and graphical human-computer interaction. There is the modeling procedure for the NPP secondary loop system simulating and modeling. In the device library there are different simulation models of secondary loop system. The secondary loop simulation model can be made up by the device simulation model with reasonable data interface. Each device simulation model could represent the practical device steady and dynamic performance with accurate calculation. The simulation results calculated in the simulation platform can represent the accuracy steady calculation and reasonable dynamic tendency of main parameters for the secondary loop system built. From the simulation platform the NPP secondary loop system simulation model can be built conveniently and fast. In the simulation platform there is the appropriate data input form and output form. The simulation platform can be used in the different purposes of simulation for NPP secondary loop as training, evaluating, operation and characteristic analysis.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"11 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":"128441041","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 conventional method for neutron-photon coupling transport calculation lacks of clear physical meanings, where the process of neutron transport and photon transport are independent, and only ensures the numbers of photons to be coupling with the neutrons. At the same time, when dealing with photoelectric effect, the nuclear data will be processed frequently, increasing the amount of calculation. By modifying the RMC codes, the deep-coupling and preprocessed photon transport is achieved. This new coupling method can satisfy the physical requirements and reduce the computational complexity while ensuring the accuracy of the calculation. At the same time, the preprocessing of the photoelectric effect nuclear data can accelerate the calculation without changing the calculation results. Through the deep-coupling and preprocessed photon transport method, the RMC codes can finished the high-precision shielding calculation. A typical LWR component is calculated with the new method, and the results prove the effectiveness.
{"title":"The Deep-Coupling and Preprocessed Photon Transport Based on RMC Codes","authors":"Pan Qingquan, Wang Kan","doi":"10.1115/ICONE26-81036","DOIUrl":"https://doi.org/10.1115/ICONE26-81036","url":null,"abstract":"The conventional method for neutron-photon coupling transport calculation lacks of clear physical meanings, where the process of neutron transport and photon transport are independent, and only ensures the numbers of photons to be coupling with the neutrons. At the same time, when dealing with photoelectric effect, the nuclear data will be processed frequently, increasing the amount of calculation. By modifying the RMC codes, the deep-coupling and preprocessed photon transport is achieved. This new coupling method can satisfy the physical requirements and reduce the computational complexity while ensuring the accuracy of the calculation. At the same time, the preprocessing of the photoelectric effect nuclear data can accelerate the calculation without changing the calculation results. Through the deep-coupling and preprocessed photon transport method, the RMC codes can finished the high-precision shielding calculation. A typical LWR component is calculated with the new method, and the results prove the effectiveness.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"208 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":"130179752","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}
Single-phase natural circulation experiments were conducted to study the flow resistance and heat transfer characteristics in a 3 × 3 rod bundle channel with the ratio of rod pitch and rod outer diameter (P/D) 1.38. The range of inlet subcooling degree is 30∼90K and the heating power is 1∼20kW. The rods are heated with constant heat flux. According to the experimental results, the flow regime under natural circulation condition is divided and the transition Reynolds number is considered as 800. The flow transition is recognized by the slope change of friction factor curve since the flow transition in the rod bundle channel is not as obvious as that in round pipe. Simultaneously, the flow transition in the rod bundle is much earlier and the upper critical Reynolds number is much larger compared to regular channel like round pipe and rectangular channel. Two correlations for laminar and transition regime are fitted to calculate the friction factor. As for the grid spacer local resistance coefficient, there is slight change at Reynolds number 800 and similarly two correlations are fitted to calculate the local resistance coefficient. The Nusselt number tendency changes at around Reynolds number 4000 but keep unchanged at transition point, which means the flow transition has no obvious effect to the heat transfer. The heat transfer results are compared with different single-phase convective heat transfer correlations. D-B and Gnielinski correlations are not suitable for the heat transfer prediction in rod bundle channel and the relative deviation is more than 20%. Weisman, Presser and Markoczy correlations predict relatively well in high Reynolds number region, and Markoczy correlation is the best of them. In low Reynolds number region, most experimental results are larger than the correlations. D-B correlation based methods may be unsuitable for the heat transfer prediction in rod bundle channel and a new correlation needs to be proposed.
{"title":"Experimental Research on the Flow Resistance and Heat Transfer Characteristics in Rod Bundle Channel","authors":"Zhiqiang Zhu, Chunping Tian, Chang-qi Yan, Jianjun Wang, Tingting Ren, Zehua Guo","doi":"10.1115/ICONE26-82195","DOIUrl":"https://doi.org/10.1115/ICONE26-82195","url":null,"abstract":"Single-phase natural circulation experiments were conducted to study the flow resistance and heat transfer characteristics in a 3 × 3 rod bundle channel with the ratio of rod pitch and rod outer diameter (P/D) 1.38. The range of inlet subcooling degree is 30∼90K and the heating power is 1∼20kW. The rods are heated with constant heat flux.\u0000 According to the experimental results, the flow regime under natural circulation condition is divided and the transition Reynolds number is considered as 800. The flow transition is recognized by the slope change of friction factor curve since the flow transition in the rod bundle channel is not as obvious as that in round pipe. Simultaneously, the flow transition in the rod bundle is much earlier and the upper critical Reynolds number is much larger compared to regular channel like round pipe and rectangular channel. Two correlations for laminar and transition regime are fitted to calculate the friction factor. As for the grid spacer local resistance coefficient, there is slight change at Reynolds number 800 and similarly two correlations are fitted to calculate the local resistance coefficient. The Nusselt number tendency changes at around Reynolds number 4000 but keep unchanged at transition point, which means the flow transition has no obvious effect to the heat transfer. The heat transfer results are compared with different single-phase convective heat transfer correlations. D-B and Gnielinski correlations are not suitable for the heat transfer prediction in rod bundle channel and the relative deviation is more than 20%. Weisman, Presser and Markoczy correlations predict relatively well in high Reynolds number region, and Markoczy correlation is the best of them. In low Reynolds number region, most experimental results are larger than the correlations. D-B correlation based methods may be unsuitable for the heat transfer prediction in rod bundle channel and a new correlation needs to be proposed.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129928162","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}
Hydrogen explosion is one of the severe threats to the integrity of containment for nuclear power plant which has drawn many experts attention to make great efforts on hydrogen related issues, espeically after the Fukushima Dai-ichi nuclear power station accident took place. However, the issue of hydrogen distribution hasn’t been closed as a result of related complex process of hydrogen transport and the particular design of each kind of facility. In the present study, CFD method has been applied to the pre-analysis on the characteristic of hydrogen mixing and stratification in a computational containment model for the sake of probable phenomena identification and instrumentaion design of experimental study in the next phase. Firstly, physical models have been verified by the experimental data from THAI HM2. Based on the determined numerical models, five typical groups of cases have been simulated, considering the effect of initial momentum, injection location, and injection direction. During the cases, only helium has been released in the vessel isothermally, on behalf of hydrogen. The results show that the backflow from the wall to the main stream in the dome and the buoyancy force may strongly dominate the helium flow, thus affacting the mixing and stratification. The eccentric injection and horizontal injection may also influence the helium distribution, in which the wall effects and rapid shifting may play important roles. However, the inference will be examined in the experiments later. All the work will be helpful for safety design and analysis of newly-built containment in China.
{"title":"Numerical Analysis on Characteristic of Hydrogen Mixing and Stratification in a Containment Model","authors":"Cheng Peng, L. Tong, Xuewu Cao","doi":"10.1115/ICONE26-81235","DOIUrl":"https://doi.org/10.1115/ICONE26-81235","url":null,"abstract":"Hydrogen explosion is one of the severe threats to the integrity of containment for nuclear power plant which has drawn many experts attention to make great efforts on hydrogen related issues, espeically after the Fukushima Dai-ichi nuclear power station accident took place. However, the issue of hydrogen distribution hasn’t been closed as a result of related complex process of hydrogen transport and the particular design of each kind of facility. In the present study, CFD method has been applied to the pre-analysis on the characteristic of hydrogen mixing and stratification in a computational containment model for the sake of probable phenomena identification and instrumentaion design of experimental study in the next phase. Firstly, physical models have been verified by the experimental data from THAI HM2. Based on the determined numerical models, five typical groups of cases have been simulated, considering the effect of initial momentum, injection location, and injection direction. During the cases, only helium has been released in the vessel isothermally, on behalf of hydrogen. The results show that the backflow from the wall to the main stream in the dome and the buoyancy force may strongly dominate the helium flow, thus affacting the mixing and stratification. The eccentric injection and horizontal injection may also influence the helium distribution, in which the wall effects and rapid shifting may play important roles. However, the inference will be examined in the experiments later. All the work will be helpful for safety design and analysis of newly-built containment in China.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"51 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":"131760682","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}
Shibao Wang, Dalin Zhang, Chenglong Wang, P. Song, J. Chen, S. Qiu, G. Su
Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k–ε model (SKE) and the realizable k–ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.
{"title":"Validation of a Code and Effect of Turbulence Model on Predicting Thermal Stratification Phenomena in the Upper Plenum of SFR","authors":"Shibao Wang, Dalin Zhang, Chenglong Wang, P. Song, J. Chen, S. Qiu, G. Su","doi":"10.1115/ICONE26-81551","DOIUrl":"https://doi.org/10.1115/ICONE26-81551","url":null,"abstract":"Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k–ε model (SKE) and the realizable k–ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"104 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131952108","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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