Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators. In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.
集成化小型核反应堆因其核安全风险低、初期投资成本低、建设周期短等特点，受到了世界各国的广泛关注。整体式小型核反应堆作为一种先进的新型核反应堆，其技术正处于探索和发展的过程中。蒸汽发生器是反应堆一次回路和二次回路之间进行能量交换的传热系统，其传热分析对反应堆的设计和开发具有重要意义。直通式蒸汽发生器(otsg)由于结构简单、换热能力强、接负荷及时等优点，是一体化小型核反应堆设计中主要采用的蒸汽发生器。RELAP5/MOD4.0是由Innovative System software, LLC开发的用于轻水反应堆(LWR)瞬态分析的商业软件。RELAP5经过多年的发展和完善，已经成为核电站各种模拟器分析计算的基础工具。然而，RELAP5在对蒸汽发生器进行建模时，只能建立与直管相关的结构模型，这对直通式蒸汽发生器的传热研究非常不便。本文以具有特定结构参数的直通式蒸汽发生器为研究对象。利用RELAP5程序对简化后的斜管模型进行了传热计算，并对螺旋管传热模型进行了理论计算。通过对给定主、次侧进口流体条件下主、次侧蒸汽出口温度、换热功率、平均换热系数和不同换热区管长进行比较，验证了RELAP5换热计算方法对于简化斜管式直流式蒸汽发生器模型的有效性。
{"title":"Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR","authors":"B. Jiang, Zhiwei Zhou, Z. Xia, Qian Sun","doi":"10.1115/icone2020-16281","DOIUrl":"https://doi.org/10.1115/icone2020-16281","url":null,"abstract":"\u0000 Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors.\u0000 RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators.\u0000 In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128333302","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 visualized experimental system is designed and constructed to investigate the bubble dynamic in a flowing liquid layer. Motivated by reducing uncertainties and digging a deep understand on the formation mechanism of boiling bubbles, the bubbles are formed by injecting air through a submerged orifice in our present work, where the influence of thermal physics, nucleation site density and dry spot are stripped. The water flow rate and the air flow rate are in the range of 72–324 ml/min and 0.8–2.0 ml/min, respectively. The bubble formation process in the smooth channel and the rib channel are investigated. The results state that increasing the liquid flow rates lead to the increasing bubble detachment frequency and the decreasing bubble detachment volume. Besides, the larger the liquid flow rate is, the closer the bubble center of mass is to the wall. The rib has a significant influence on the bubble formation process. In the rib channel, it is more difficult for bubbles to detach from the orifice compared that in a smooth channel. Besides, the bubble detachment volume in a rib channel is larger than it in a smooth channel.
{"title":"Experimental Study of Bubble Behaviour in a Flowing Liquid Layer","authors":"Zhengzheng Zhang, Liangxing Li, Shuanglei Zhang, Afnan Saleem","doi":"10.1115/icone2020-16397","DOIUrl":"https://doi.org/10.1115/icone2020-16397","url":null,"abstract":"\u0000 A visualized experimental system is designed and constructed to investigate the bubble dynamic in a flowing liquid layer. Motivated by reducing uncertainties and digging a deep understand on the formation mechanism of boiling bubbles, the bubbles are formed by injecting air through a submerged orifice in our present work, where the influence of thermal physics, nucleation site density and dry spot are stripped. The water flow rate and the air flow rate are in the range of 72–324 ml/min and 0.8–2.0 ml/min, respectively. The bubble formation process in the smooth channel and the rib channel are investigated. The results state that increasing the liquid flow rates lead to the increasing bubble detachment frequency and the decreasing bubble detachment volume. Besides, the larger the liquid flow rate is, the closer the bubble center of mass is to the wall. The rib has a significant influence on the bubble formation process. In the rib channel, it is more difficult for bubbles to detach from the orifice compared that in a smooth channel. Besides, the bubble detachment volume in a rib channel is larger than it in a smooth channel.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"66 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124716188","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 recent years, hybrid nanofluids, as a new kind of working fluid, have been widely studied because they possessing better heat transfer performance than single component nanofluids when prepared with proper constituents and proportions. The application of hybrid nanofluids in nuclear power system as a working fluid is an effective way of improving the capability of In-Vessel Retention (IVR) when the reactor is in a severe accident. In order to obtain hybrid nanofluids with excellent heat transfer performance, three kinds of hybrid nanofluids with high thermal conductivity are measured by transient plane source method, and their viscosity and stability are also investigated experimentally. These experimental results are used to evaluate the heat transfer efficiency of hybrid nanofluids. The results show that: (1) The thermal conductivity of hybrid nanofluids increases with increasing temperature and volume concentration. When compared to the base fluid, the thermal conductivity of Al2O3-CuO/H2O, Al2O3-C/H2O and AlN-TiO2/H2O nanofluids at 0.25% volume concentration increased by 36%, 24%, and 22%, respectively. (2) Surfactants can improve the stability of hybrid nanofluids. The Zeta potential value is related to the thermal conductivity of the hybrid nanofluids, and it could be used to explain the relationship between the thermal conductivity of the hybrid nanofluids and the dispersion. It also could provide a reference for subsequent screening of high thermal conductivity nanofluids. (3) The addition of C/H2O can effectively reduce the dynamic viscosity coefficient of hybrid nanofluids. (4) The analysis of heat transfer efficiency of the hybrid nanofluids found that both Al2O3-CuO/H2O and Al2O3-C/H2O have better heat transfer ability than water under certain mixing conditions. This study is conducive to further optimizing hybrid nanofluids and its application to the In-Vessel Retention in severe reactor accidents.
{"title":"Experimental Study on the Thermal Properties and Stability of Hybrid Nanofluids and Evaluation of its Heat Exchange Efficiency","authors":"Yubai Xiao, Hu Zhang, Junmei Wu","doi":"10.1115/icone2020-16630","DOIUrl":"https://doi.org/10.1115/icone2020-16630","url":null,"abstract":"\u0000 In recent years, hybrid nanofluids, as a new kind of working fluid, have been widely studied because they possessing better heat transfer performance than single component nanofluids when prepared with proper constituents and proportions. The application of hybrid nanofluids in nuclear power system as a working fluid is an effective way of improving the capability of In-Vessel Retention (IVR) when the reactor is in a severe accident. In order to obtain hybrid nanofluids with excellent heat transfer performance, three kinds of hybrid nanofluids with high thermal conductivity are measured by transient plane source method, and their viscosity and stability are also investigated experimentally. These experimental results are used to evaluate the heat transfer efficiency of hybrid nanofluids. The results show that: (1) The thermal conductivity of hybrid nanofluids increases with increasing temperature and volume concentration. When compared to the base fluid, the thermal conductivity of Al2O3-CuO/H2O, Al2O3-C/H2O and AlN-TiO2/H2O nanofluids at 0.25% volume concentration increased by 36%, 24%, and 22%, respectively. (2) Surfactants can improve the stability of hybrid nanofluids. The Zeta potential value is related to the thermal conductivity of the hybrid nanofluids, and it could be used to explain the relationship between the thermal conductivity of the hybrid nanofluids and the dispersion. It also could provide a reference for subsequent screening of high thermal conductivity nanofluids. (3) The addition of C/H2O can effectively reduce the dynamic viscosity coefficient of hybrid nanofluids. (4) The analysis of heat transfer efficiency of the hybrid nanofluids found that both Al2O3-CuO/H2O and Al2O3-C/H2O have better heat transfer ability than water under certain mixing conditions. This study is conducive to further optimizing hybrid nanofluids and its application to the In-Vessel Retention in severe reactor accidents.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"145 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123428655","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In order to meet the demand of continuous innovation of technologies and the general trend of autonomous nuclear power plants design and export of nuclear power plants, it is necessary to develop an autonomous LOCA analysis platform and corresponding analysis methods for the most complex design basis accidents. In this paper, the characteristics of LOCA analysis platform ARSAC, designed by Nuclear Power Institute of China, and the code ARSAC-K which meets the requirements of the US Federal Code 10 CFR 50.46 Appendix K model are introduced as well as a set of LOCA analysis methods and modeling methods. Based on the international advanced LOCA analysis code development concept, the code ARSAC has made new breakthroughs in matrix algorithms, key thermal hydraulic models and so on. Validation work has also been carried out in-depth. A set of advanced LOCA analysis methods has been developed using code ARSAC-K and advanced power plant parameter sampling methods. Analysis on LBLOCA of nuclear power plants with code ARSAC-K was performed, and the impact of different modeling methods on the LOCA analysis results was studied. To ensure the rationality and conservativeness of the analysis results, a set of reasonable and conservative modeling methods is fixed on the basis of a large number of sensitivity analyses for subsequent analysis and calculation. In the future, a lot of optimization work will be done to improve the LOCA code and corresponding methods.
{"title":"Research on Analysis and Modeling Methods for LBLOCA in Nuclear Power Plants Based on Autonomous LOCA Analysis Platform ARSAC","authors":"Jiayue Zhou, Dan Wu, Shuhua Ding, G.-L. Jiang","doi":"10.1115/icone2020-16820","DOIUrl":"https://doi.org/10.1115/icone2020-16820","url":null,"abstract":"\u0000 In order to meet the demand of continuous innovation of technologies and the general trend of autonomous nuclear power plants design and export of nuclear power plants, it is necessary to develop an autonomous LOCA analysis platform and corresponding analysis methods for the most complex design basis accidents. In this paper, the characteristics of LOCA analysis platform ARSAC, designed by Nuclear Power Institute of China, and the code ARSAC-K which meets the requirements of the US Federal Code 10 CFR 50.46 Appendix K model are introduced as well as a set of LOCA analysis methods and modeling methods. Based on the international advanced LOCA analysis code development concept, the code ARSAC has made new breakthroughs in matrix algorithms, key thermal hydraulic models and so on. Validation work has also been carried out in-depth. A set of advanced LOCA analysis methods has been developed using code ARSAC-K and advanced power plant parameter sampling methods. Analysis on LBLOCA of nuclear power plants with code ARSAC-K was performed, and the impact of different modeling methods on the LOCA analysis results was studied. To ensure the rationality and conservativeness of the analysis results, a set of reasonable and conservative modeling methods is fixed on the basis of a large number of sensitivity analyses for subsequent analysis and calculation. In the future, a lot of optimization work will be done to improve the LOCA code and corresponding methods.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"83 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132559914","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 water separator is the core equipment of nuclear power plant. It is very vital for improving the efficiency of the steam separator to study the impact characteristics between the droplets and the curved dry wall of the steam separator under the action of the rotating air flow. In this paper, the characteristics of droplet impinging on the dry wall in the rotating flow field are analyzed by establishing a visualization experimental device. A high-speed camera was used to capture the impact of droplets with different diameters on the dry wall of a non-wetting curved surface at different gas velocities. At the same time, using image processing tool in MATLAB to obtain image boundary information. The characteristics of spreading factor, droplet deformation factor and initial diffusion velocity of droplets impacting the surface dry wall under different wind speeds are studied.
{"title":"Analysis of the Behavior of Droplet Impinging on a Curved Dry Wall in a Rotating Flow Field","authors":"S. Ouyang, Z. Xiong","doi":"10.1115/icone2020-16404","DOIUrl":"https://doi.org/10.1115/icone2020-16404","url":null,"abstract":"\u0000 Steam water separator is the core equipment of nuclear power plant. It is very vital for improving the efficiency of the steam separator to study the impact characteristics between the droplets and the curved dry wall of the steam separator under the action of the rotating air flow. In this paper, the characteristics of droplet impinging on the dry wall in the rotating flow field are analyzed by establishing a visualization experimental device. A high-speed camera was used to capture the impact of droplets with different diameters on the dry wall of a non-wetting curved surface at different gas velocities. At the same time, using image processing tool in MATLAB to obtain image boundary information. The characteristics of spreading factor, droplet deformation factor and initial diffusion velocity of droplets impacting the surface dry wall under different wind speeds are studied.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"44 9","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132654511","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 outlines a system level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The protected loss of flow (LOF) accident was selected as an exemplified design basis accident to demonstrate the analysis procedure. The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor was adopted as a research reactor model in the study. Two system level dynamics codes, RELAP5-3D and PARET, were employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the system code models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios. The canonical transient predictions on the LOF accident yielded from the two system codes mentioned above have demonstrated some noticeable yet acceptable discrepancies. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the RELAP5-3D model and the data analytic engines provided by the RAVEN framework developed by INL. The sensitivity information reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions delivered by the framework.
本文概述了一个包含灵敏度和不确定度成分的研究堆系统级安全分析程序。并以保护失流(LOF)事故为例，对分析过程进行了说明。本研究采用概念型NIST (National Institute of Standards and Technology)水平裂芯研究堆作为研究堆模型。本文采用RELAP5-3D和PARET两种系统级动力学代码进行对比研究。本工作的主要目的是证明除了传统的系统代码模型预测之外，集成灵敏度和不确定性信息的分析能力，用于研究反应堆在事故瞬态情景下的热工水力(T/H)安全特性。从上述两种系统代码得到的LOF事故的典型瞬态预测显示出一些明显但可接受的差异。为了更好地理解模拟中观察到的差异，通过将RELAP5-3D模型与INL开发的RAVEN框架提供的数据分析引擎耦合进行敏感性和不确定性分析。灵敏度信息揭示了在正常运行和设计基础瞬态中，包层峰值温度随不同边界和初始参数变化等关键参数的意义。不确定性分析告知了由各种输入分量的误差引起的响应偏差。敏感性和不确定性信息都将被纳入安全分析框架，作为框架提供的安全特性预测的一部分。
{"title":"Sensitivity and Uncertainty Information Incorporated Loss of Flow Accident Analyses for Research Reactors","authors":"Tao Liu, Zeyun Wu","doi":"10.1115/icone2020-16242","DOIUrl":"https://doi.org/10.1115/icone2020-16242","url":null,"abstract":"\u0000 This paper outlines a system level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The protected loss of flow (LOF) accident was selected as an exemplified design basis accident to demonstrate the analysis procedure. The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor was adopted as a research reactor model in the study. Two system level dynamics codes, RELAP5-3D and PARET, were employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the system code models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios. The canonical transient predictions on the LOF accident yielded from the two system codes mentioned above have demonstrated some noticeable yet acceptable discrepancies. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the RELAP5-3D model and the data analytic engines provided by the RAVEN framework developed by INL. The sensitivity information reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions delivered by the framework.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"86 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133046907","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}
L. Peiying, Deng Jian, Z. Lei, Qian Libo, Cai Rong, Ma Yugao, Yu Hongxing
With liquid metal like lead-bismuth alloy (LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt = A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism.
{"title":"Investigation of Turbulent Prt Model and the Segmentation Rule for LBE Turbulent Heat Transfer","authors":"L. Peiying, Deng Jian, Z. Lei, Qian Libo, Cai Rong, Ma Yugao, Yu Hongxing","doi":"10.1115/icone2020-16724","DOIUrl":"https://doi.org/10.1115/icone2020-16724","url":null,"abstract":"\u0000 With liquid metal like lead-bismuth alloy (LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt = A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"145 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127255906","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}
Most of today’s operating nuclear plants that were originally designed for 30 or 40-year life are facing the long-term operation issues. Therefore, it is of meaningful importance to assess the time-dependent degradation and the ageing of the relevant nuclear systems, structures, and components (SSCs) because of resulting loss of structural capacity. In this framework, the inverse method is implemented starting from temperatures at an accessible boundary, which are measured through a monitoring system. The reconstruction technique uses the elaborated signal provided by the monitoring system to determine temperature at inaccessible surface: this is the so-called inverse heat transfer problem (IHTP). The inverse space marching method is applied. Analytical and numerical studies are performed taking into account thermal transient conditions in order to determine thermal loads. In particular, the developed code demonstrates to be able to reconstruct temperature and stress profiles in any section of the pipe with a good accuracy. In addition, the thermal loads obtained suggest that the investigated transient condition is not able to jeopardise the integrity of NPP, confirming the possibility of the plant extension of life.
{"title":"Inverse Heat Conduction Problem in Estimating NPP Pipeline Performance","authors":"S. A. Cancemi, R. L. Frano","doi":"10.1115/icone2020-16129","DOIUrl":"https://doi.org/10.1115/icone2020-16129","url":null,"abstract":"\u0000 Most of today’s operating nuclear plants that were originally designed for 30 or 40-year life are facing the long-term operation issues. Therefore, it is of meaningful importance to assess the time-dependent degradation and the ageing of the relevant nuclear systems, structures, and components (SSCs) because of resulting loss of structural capacity.\u0000 In this framework, the inverse method is implemented starting from temperatures at an accessible boundary, which are measured through a monitoring system. The reconstruction technique uses the elaborated signal provided by the monitoring system to determine temperature at inaccessible surface: this is the so-called inverse heat transfer problem (IHTP).\u0000 The inverse space marching method is applied.\u0000 Analytical and numerical studies are performed taking into account thermal transient conditions in order to determine thermal loads. In particular, the developed code demonstrates to be able to reconstruct temperature and stress profiles in any section of the pipe with a good accuracy.\u0000 In addition, the thermal loads obtained suggest that the investigated transient condition is not able to jeopardise the integrity of NPP, confirming the possibility of the plant extension of life.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"2011 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127369872","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}
Panpan Wen, Gen Li, Jinchen Gao, Yupeng Li, A. Yamaji, Junjie Yan
The collision dynamics between two droplets plays an important role in various disciplines of nature and practical interests, such as fuel-coolant interaction (FCI), fuel combustion in engines, and various spraying process. FCI presents in nuclear reactor severe accident when the melt relocates into the coolant in the lower head with violent disturbance and vigorous heat transfer. The purpose of this study is to investigate the collision behavior of melt droplets during fuel-coolant interaction. The collision of two equal-sized droplets has been simulated in 3D by using the volume of fluid (VOF) and adaptive mesh refinement method. The numerical simulations of tetradecane droplet collision were carried out to validate the numerical methods. The results showed good agreement with the experiments. Furthermore, the simulations of uranium dioxide (UO2) droplets collision in coolant were carried out. The results showed that the contact area between droplets and coolant increased with time first and then decreased. With the increase of Weber number, the contact area of maximum in the droplet collision increased. Break happened in the later period and many child droplets formed. The number of child droplets increased with the increase of Weber number. In addition, the size distribution of little droplets was investigated.
{"title":"Numerical Study of Collision Behavior of Melt Drops During Fuel-Coolant Interaction","authors":"Panpan Wen, Gen Li, Jinchen Gao, Yupeng Li, A. Yamaji, Junjie Yan","doi":"10.1115/icone2020-16206","DOIUrl":"https://doi.org/10.1115/icone2020-16206","url":null,"abstract":"\u0000 The collision dynamics between two droplets plays an important role in various disciplines of nature and practical interests, such as fuel-coolant interaction (FCI), fuel combustion in engines, and various spraying process. FCI presents in nuclear reactor severe accident when the melt relocates into the coolant in the lower head with violent disturbance and vigorous heat transfer. The purpose of this study is to investigate the collision behavior of melt droplets during fuel-coolant interaction. The collision of two equal-sized droplets has been simulated in 3D by using the volume of fluid (VOF) and adaptive mesh refinement method. The numerical simulations of tetradecane droplet collision were carried out to validate the numerical methods. The results showed good agreement with the experiments. Furthermore, the simulations of uranium dioxide (UO2) droplets collision in coolant were carried out. The results showed that the contact area between droplets and coolant increased with time first and then decreased. With the increase of Weber number, the contact area of maximum in the droplet collision increased. Break happened in the later period and many child droplets formed. The number of child droplets increased with the increase of Weber number. In addition, the size distribution of little droplets was investigated.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"52 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129047189","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}
Xu Han, Xiuzhong Shen, Toshihiro Yamamoto, K. Nakajima, T. Hibiki
In the present paper, the local two-phase flow parameters were measured with a four-sensor optical probe in adiabatic upward air-water two-phase flows in a 6 × 6 vertical rod bundle with rod diameter of 10 mm, pitch of 16.7 mm, square channel box side length of 100 mm and hydraulic equivalent diameter (DH) of 18.7 mm. The local measurements were performed in an octant triangular region of the rod bundle cross-section at the axial position with height-to-diameter ratio (z/DH) of 149 under a total of 16 flow conditions. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity were obtained in the four-sensor probe measurements. Both of the measured void fraction and IAC show their radial local distributions with the core-peaking and wall-peaking shapes which are closely linked with the superficial velocity of two phases. The distribution shapes tend to change from the core-peaking to the wall-peaking when the superficial liquid velocity () increases and the superficial gas velocity () decreases. The measured diameters of local bubbles keep the similar values in the measuring cross-section, increase when the increases and decrease when the increases. The bubbles behave with the velocities whose main flow direction component keeps a typical radial power-law (core-peaking) profile and whose cross-sectional velocity components show a significant trend of bubbles migrating from the center to the wall region of the channel box especially under high conditions. The area-averaged results of void fraction and IAC were obtained by a cross-sectional area-averaging scheme. The resultant area-averaged void fraction and IAC were used to check the void fraction predicting capability of two drift-flux correlations and the IAC prediction performance of two IAC correlations respectively. The applicability of these correlations to the rod bundle geometry was discussed and concluded finally in this paper.
{"title":"Flow Characteristics of Upward Two-Phase Flows in a Rod Bundle Geometry","authors":"Xu Han, Xiuzhong Shen, Toshihiro Yamamoto, K. Nakajima, T. Hibiki","doi":"10.1115/icone2020-16740","DOIUrl":"https://doi.org/10.1115/icone2020-16740","url":null,"abstract":"\u0000 In the present paper, the local two-phase flow parameters were measured with a four-sensor optical probe in adiabatic upward air-water two-phase flows in a 6 × 6 vertical rod bundle with rod diameter of 10 mm, pitch of 16.7 mm, square channel box side length of 100 mm and hydraulic equivalent diameter (DH) of 18.7 mm. The local measurements were performed in an octant triangular region of the rod bundle cross-section at the axial position with height-to-diameter ratio (z/DH) of 149 under a total of 16 flow conditions. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity were obtained in the four-sensor probe measurements. Both of the measured void fraction and IAC show their radial local distributions with the core-peaking and wall-peaking shapes which are closely linked with the superficial velocity of two phases. The distribution shapes tend to change from the core-peaking to the wall-peaking when the superficial liquid velocity (<jf>) increases and the superficial gas velocity (<jg>) decreases. The measured diameters of local bubbles keep the similar values in the measuring cross-section, increase when the <jg> increases and decrease when the <jf> increases. The bubbles behave with the velocities whose main flow direction component keeps a typical radial power-law (core-peaking) profile and whose cross-sectional velocity components show a significant trend of bubbles migrating from the center to the wall region of the channel box especially under high <jf> conditions. The area-averaged results of void fraction and IAC were obtained by a cross-sectional area-averaging scheme. The resultant area-averaged void fraction and IAC were used to check the void fraction predicting capability of two drift-flux correlations and the IAC prediction performance of two IAC correlations respectively. The applicability of these correlations to the rod bundle geometry was discussed and concluded finally in this paper.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"8 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131686748","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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