International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第6页

Uncertainty analysis of molten corium In-Vessel Retention under External Reactor Vessel Cooling 反应堆外容器冷却条件下容器内熔铈滞留的不确定性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-12-01 DOI: 10.1016/j.jandt.2024.04.002

Shihao Wu, Dong Wang, Yapei Zhang, Yukun Zhou, Wenxi Tian, Suizheng Qiu, G.H. Su

引用次数: 0

Conceptual design of a novel megawatt portable nuclear power system by supercritical carbon dioxide brayton cycle coupled with heat pipe cooled reactor 超临界二氧化碳布雷顿循环与热管冷却反应堆新型兆瓦级便携式核能系统的概念设计

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-12-01 DOI: 10.1016/j.jandt.2024.04.003

Simao Guo, Xiaoyu Guo, Guanbo Wang, Yuchuan Guo, Zi Wang, Jun Leng, Wankui Yang, Bin Tang, Yaoguang Liu, Dazhi Qian

{"title":"Conceptual design of a novel megawatt portable nuclear power system by supercritical carbon dioxide brayton cycle coupled with heat pipe cooled reactor","authors":"Simao Guo, Xiaoyu Guo, Guanbo Wang, Yuchuan Guo, Zi Wang, Jun Leng, Wankui Yang, Bin Tang, Yaoguang Liu, Dazhi Qian","doi":"10.1016/j.jandt.2024.04.003","DOIUrl":"10.1016/j.jandt.2024.04.003","url":null,"abstract":"<div><div>In order to meet realistic requirements for flexible and reliable power supply in special scenarios such as remote areas and deep ocean exploration, a novel megawatt portable nuclear power system called SUPERHERO (SUPERcritical carbon dioxide cycle-HEat pipe ReactOr power system) is proposed in this paper. The system consists of a heat pipe-cooled reactor and a closed supercritical carbon dioxide (S–CO<sub>2</sub>) Brayton power cycle. Different cycle concepts were compared, and the simple recuperated cycle was adopted for compactness and simplicity of the system. The multi-objective optimization is performed to maximize cycle efficiency and minimize system volume, resulting in optimized parameters for a 1MWe system. By utilizing uranium nitride (UN) modular fuel elements and high heat transfer capacity potassium heat pipes, a compact reactor scheme with 3.13MWt is determined that can operate for 15 years without refueling. The neutronics and thermal analysis of the reactor are carried out, and the results show that the reactor has safety features such as a low peak factor, negative temperature feedback coefficient, sufficient reactivity compensation and control, and anti-failure of single heat pipe. The primary heat exchanger is also designed using the CFD method, and the heat exchange tubes combined with ribbed plates and self-supporting structures are used to enhance heat transfer and ensure structural stability. All these studies outline the general features of SUPERHERO, demonstrating that SUPERHERO has great development potential.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 4","pages":"Pages 168-188"},"PeriodicalIF":0.0,"publicationDate":"2023-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141025153","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Experimental study of the interaction between molten corium and new sacrificial material for ex-vessel melt retention 熔融铈与用于出容器熔体保持的新型牺牲材料之间相互作用的实验研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-12-01 DOI: 10.1016/j.jandt.2024.04.001

Nan Li , Erhui Chen , Li Zhang , Hongliang Wang , Yidan Yuan , Bo Li

引用次数: 0

Research on characteristic of flow instability in a two-phase natural circulation loop with parallel once-through steam generators 并联直通式蒸汽发生器两相自然循环回路流动不稳定特性研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-12-01 DOI: 10.1016/j.jandt.2024.05.003

Ang Li, Yuqing Chen, Yuxian Rao, Kehan Ouyang

{"title":"Research on characteristic of flow instability in a two-phase natural circulation loop with parallel once-through steam generators","authors":"Ang Li, Yuqing Chen, Yuxian Rao, Kehan Ouyang","doi":"10.1016/j.jandt.2024.05.003","DOIUrl":"10.1016/j.jandt.2024.05.003","url":null,"abstract":"<div><div>When there are two sets of parallel once through-steam generators(OTSG) in the natural circulation loop, the thermal-hydraulic characteristics are complicated. Flow instability often occurs in such loops. In order to alleviate the harm caused by flow instability and enhance the natural circulation capacity, it is necessary to study the characteristics of flow instability in depth. In this paper, a thermal-hydraulic computational model is developed for a two-phase natural circulation loop with two parallel OTSGs based on the RELAP5 program. And the model is validated. The thermal-hydraulic parameters of the loop were calculated for different decay heat conditions. The flow instability of the loop is found to be characterized by a clear division into three zones in the decay heat interval. The concept of “load-flow ratio” is proposed to analyze the critical point of flow instability under different operating conditions. The reason for the flow instability zones of the loop is explained. The conclusions of the relevant research can provide a basis for an in-depth study of the characteristics of flow instability and the development of operational schemes to mitigate flow instability.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 4","pages":"Pages 189-199"},"PeriodicalIF":0.0,"publicationDate":"2023-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143276784","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research on thermal stratification of surge line in the marine reactor based on fluid-solid thermal coupling model 基于流固热耦合模型的海洋反应堆涌浪线热分层研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-12-01 DOI: 10.1016/j.jandt.2024.05.002

Jiao Meng, Xinwen Zhao, Yongfa Zhang

{"title":"Research on thermal stratification of surge line in the marine reactor based on fluid-solid thermal coupling model","authors":"Jiao Meng, Xinwen Zhao, Yongfa Zhang","doi":"10.1016/j.jandt.2024.05.002","DOIUrl":"10.1016/j.jandt.2024.05.002","url":null,"abstract":"<div><div>Due to the requirements of ship movement, the power of the marine reactor is constantly changing, and the expansion and contraction of the primary coolant make the fluid flow in and out through the surge line. This paper establishes a fluid-solid thermal coupling model for the thermal stratification of a Marine reactor surge line. The low Reynolds number eddy viscosity model and the curvature modification of the bending pipe are used in this model. Qiao experiment is provided to verify the model's applicability to the thermal stratification analysis. The typical working conditions of the Marine reactor are analyzed, and the results show that: the temperature difference between the top and bottom of the pipe changes constantly with the fluctuation of the fluid, the maximum temperature difference in the horizontal section is 59.6 K, and the maximum temperature difference in the vertical area is 23.9 K. The maximum stress fluctuation of the horizontal section is 36 MPa, while that of the vertical section is 5 MPa. In the non-destructive testing, many small pits were found in the horizontal section of the pipeline, while no obvious damage was found on the surface of the vertical section, which is consistent with the results of our numerical analysis.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 4","pages":"Pages 207-217"},"PeriodicalIF":0.0,"publicationDate":"2023-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141049471","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Excitation function of neutron induced reaction on isotopes of cadmium (106Cd, 108Cd &110Cd) in the energy range of 13–15 MeV 13-15 MeV 能量范围内镉同位素（106Cd、108Cd 和 110Cd）的中子诱导反应激发函数

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-09-01 DOI: 10.1016/j.jandt.2023.12.001

Amsalu Benti Ejerso , Habtamu Fekadu Etefa , Francis Birhanu Dejene

{"title":"Excitation function of neutron induced reaction on isotopes of cadmium (106Cd, 108Cd &110Cd) in the energy range of 13–15 MeV","authors":"Amsalu Benti Ejerso , Habtamu Fekadu Etefa , Francis Birhanu Dejene","doi":"10.1016/j.jandt.2023.12.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.12.001","url":null,"abstract":"<div><p>This study focuses on the theoretical calculations of neutron-induced reaction cross-sections on several stable isotopes of cadmium within the energy range of 13–15 MeV. The nuclei produced in this process include <sup>105</sup>Cd, <sup>106</sup>Ag, <sup>107</sup>Cd, <sup>109</sup>Cd, and <sup>110</sup>Ag. To calculate the reaction cross-sections for specific reactions such as <sup>106</sup>Cd (n, 2n) <sup>105</sup>Cd, <sup>106</sup>Cd (n, p) <sup>106</sup>Ag, <sup>108</sup>Cd (n, 2n) <sup>107</sup>Cd, <sup>110</sup>Cd (n, 2n) <sup>109</sup>Cd, and <sup>110</sup>Cd (n, p) <sup>110</sup>Ag, we employed the Alice 91-based computer code. The calculated results were then compared and tabulated alongside experimental data obtained from the International Atomic Energy Agency (IAEA) data source, EXFOR library. Additionally, we plotted the reaction cross-sections against neutron energy (excitation function) for each reaction channel. Our findings demonstrate a strong agreement between the calculated and experimental data within the specified neutron energy range. Furthermore, we introduced several statistical parameters to assess the fitting quality between the theoretically calculated and experimentally measured reaction cross-section values.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 3","pages":"Pages 137-143"},"PeriodicalIF":0.0,"publicationDate":"2023-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605023000534/pdfft?md5=16ea982a2662716676be54d9a068b5f4&pid=1-s2.0-S2468605023000534-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139099558","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Numerical simulation and structural optimization of spiral finned tube thermal energy storage 螺旋翅片管热能存储的数值模拟和结构优化

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-09-01 DOI: 10.1016/j.jandt.2023.12.002

Yunfei Ma , Yu Meng , Jingyu Li , Weixiong Chen , Xiaohu Yang , Shaodan Li , Daotong Chong , Junjie Yan

{"title":"Numerical simulation and structural optimization of spiral finned tube thermal energy storage","authors":"Yunfei Ma , Yu Meng , Jingyu Li , Weixiong Chen , Xiaohu Yang , Shaodan Li , Daotong Chong , Junjie Yan","doi":"10.1016/j.jandt.2023.12.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.12.002","url":null,"abstract":"<div><p>Thermal energy storage (TES) has emerged as a promising solution to enhance nuclear safety by passively removing decay heat during reactor shutdown and accidents, thus preventing overheating of the reactor core and protecting the integrity of containment barriers. The research of TES with different structures has broad application prospects and practical significance. In this study, the melting process of phase change material (PCM) in various TES structures was simulated by the numerical simulation method in two and three dimensions. The mechanism of gravity-driven natural convection enhancing heat transfer was revealed. The effects of different TES structures, fin pitch, fin height, and fin thickness on heat transfer performance were studied. The results showed that many vortices formed by liquid PCM are the main reason for enhancing the natural convection. Adding fins could greatly accelerate the heat storage process. The melting time of PCM in annular and spiral finned tube TES was 47.3% and 61.3% less than that in smooth tube TES, respectively. In the present study, the heat transfer effect was enhanced as the spiral fin pitch became small, the fin height increased and the fin thickness increased. Two opposite effects of fin structure on the natural convection were revealed: (1) positive effect provided by heat transfer enhancement and (2) negative effect produced by blockage.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 3","pages":"Pages 123-136"},"PeriodicalIF":0.0,"publicationDate":"2023-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605023000546/pdfft?md5=af5eab7762c7bda526e3471fbee64dd6&pid=1-s2.0-S2468605023000546-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139099557","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility 基于ECC热混合试验装置的热液压系统代码LOCUST V1.2分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-09-01 DOI: 10.1016/j.jandt.2023.10.001

Wen Ding , Kui Zhang , Dalin Zhang , Ronghua Chen , Zhongyun Ju , Caihong Xu , Ting Wang , Wenxi Tian , Suizheng Qiu

{"title":"Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility","authors":"Wen Ding , Kui Zhang , Dalin Zhang , Ronghua Chen , Zhongyun Ju , Caihong Xu , Ting Wang , Wenxi Tian , Suizheng Qiu","doi":"10.1016/j.jandt.2023.10.001","DOIUrl":"10.1016/j.jandt.2023.10.001","url":null,"abstract":"<div><p>The safety injection (SI) system might be put into use, and the coolant at high temperature will mix with the supcooled water under Loss Of Coolant Accident (LOCA). The thermo-hydraulic phenomena associated with the thermal mixing of coolant and supcooled water will directly affect the judgment on core reflooding. China General Nuclear Power Group (CGN) independently developed and designed a thermal hydraulic system code named LOCUST, and the verification of thermal mixing models in LOCUST is necessary. This paper will introduce the thermal mixing tests at the T-junction based on Emergency Core Cooling System in Xi'an Jiaotong University (ECCS-XJTU) experimental facility, which were mainly conducted for the mixing between subcooled water injected from the SI pipe with a range of 25–125 kg/h in mass flow and the pure steam in the primary pipe with a range of 100–500 kg/h in mass flow. Besides, the fluid temperature and the dynamic vapor quality after mixing in tests are analyzed, and the simulations of 25 thermal mixing tests using the LOCUST 1.2 code are performed. The results show that the maximum relative error of LOCUST in mass flow of liquid is within 13.8 %, and the maximum relative error of LOCUST in temperature is within 8 %, which validates the reliability and accuracy of simulations of LOCUST for two-phase thermal mixing in LOCA.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 3","pages":"Pages 115-122"},"PeriodicalIF":0.0,"publicationDate":"2023-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605023000522/pdfft?md5=a1b5de1ccb20676d3837619dce7d336e&pid=1-s2.0-S2468605023000522-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135455315","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Advanced manufacturing technologies for enhancing security in nuclear and radiological materials transport 加强核材料和放射性材料运输安全的先进制造技术

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-09-01 DOI: 10.1016/j.jandt.2024.01.001

Kunal Mondal, Ryan Karkkainen, Oscar Martinez, Isaac Sikkema, Mahim Mathur, Sam Hollifield, Mingyan Li

{"title":"Advanced manufacturing technologies for enhancing security in nuclear and radiological materials transport","authors":"Kunal Mondal, Ryan Karkkainen, Oscar Martinez, Isaac Sikkema, Mahim Mathur, Sam Hollifield, Mingyan Li","doi":"10.1016/j.jandt.2024.01.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2024.01.001","url":null,"abstract":"<div><p>Advanced manufacturing technologies have transformed various industries, including nuclear security areas such as nuclear and radiological transport. This review manuscript describes the intersection of advanced manufacturing technologies and their applications to enhance the safety, efficiency, and reliability of nuclear and radiological transport. The manuscript discusses key technologies such as additive manufacturing, advanced materials, robotics, and data analytics, highlighting their aids to consolidate nuclear security in the background of nuclear and radiological transport. By providing a comprehensive perspective overview of these advancements, this review provides a deeper understanding of the potential benefits and challenges related to the integration of advanced manufacturing into this critical sector.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 3","pages":"Pages 144-150"},"PeriodicalIF":0.0,"publicationDate":"2023-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000012/pdfft?md5=7ea204885ed24adb3a8326be2e7f4a69&pid=1-s2.0-S2468605024000012-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139434013","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research on bubble generation method based on flow control mechanism 基于流动控制机理的气泡生成方法研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-08-01 DOI: 10.1016/j.jandt.2023.08.003

J. Yin, Wuguang Chen, Yu Song, Dezhong Wang

引用次数: 0