Progress in Nuclear Energy最新文献_第5页

Design and optimization of the components of a molten salt, thorium-fueled accelerator driven system 熔盐钍燃料加速器驱动系统组件的设计与优化

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-15 DOI: 10.1016/j.pnucene.2024.105486

Faisal Aldebie , Rodolfo Vaghetto , Yassin Hassan

{"title":"Design and optimization of the components of a molten salt, thorium-fueled accelerator driven system","authors":"Faisal Aldebie , Rodolfo Vaghetto , Yassin Hassan","doi":"10.1016/j.pnucene.2024.105486","DOIUrl":"10.1016/j.pnucene.2024.105486","url":null,"abstract":"<div><div>Molten Salt Accelerator-Driven Systems (MoSTADS) have been attracting a lot of research interest lately due to their unique characteristics and advantages, including reduced radiation damage of the fuel, and stable operation achieved through online fuel feeding process. Simulations of an experimental molten salt test facility being developed in the Thermal Hydraulic Research Laboratory (THRL) at Texas A&M University, were conducted using Monte Carlo radiation transport methods, to design and optimize selected components of the system. The system consists of a proton beam generated by an accelerator, impinging on a target to generate neutrons, which can be used induce fission reactions within a thorium fueled, high-temperature molten salt forced convection test loop. Parametric studies were performed to optimize several key components of the system including target material, proton beam energy, target thickness and location, and reflector thickness. Furthermore, in order to ensure the safe operation of the facility, parametric studies were also performed to identify the composition and thickness of the system shielding that would be needed to satisfy acceptable exposure limits.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105486"},"PeriodicalIF":3.3,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142434142","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Modeling and benchmarking XRF analysis using MCNP for applications in accident tolerant fuel and cladding 利用 MCNP 对 XRF 分析进行建模和基准测试，以应用于耐事故燃料和包壳

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-14 DOI: 10.1016/j.pnucene.2024.105487

James Cahill, Victoria Davis, Caleb King, Lilith Miller, Tristan Norrgard, Carlos E. Castano, Reza Mohammadi, Jessika Rojas, Braden Goddard

{"title":"Modeling and benchmarking XRF analysis using MCNP for applications in accident tolerant fuel and cladding","authors":"James Cahill, Victoria Davis, Caleb King, Lilith Miller, Tristan Norrgard, Carlos E. Castano, Reza Mohammadi, Jessika Rojas, Braden Goddard","doi":"10.1016/j.pnucene.2024.105487","DOIUrl":"10.1016/j.pnucene.2024.105487","url":null,"abstract":"<div><div>There is an interest in using nondestructive testing (NDT) methods for the preliminary investigation of accident-tolerant fuel cladding materials, such as chromium (Cr) coated Zircaloy-4 (Zr4). One promising application is X-ray fluorescence (XRF) analysis. Computational methods, such as Monte Carlo N-Particle Transport (MCNP) 6.2, can be used to expand algorithms based on XRF measurements, however, it has been demonstrated that MCNP is more sensitive to modeling imperfections at lower energies (<span><math><mrow><mo>≤</mo></mrow></math></span> 80 keV). In this work, several MCNP models were developed to evaluate the XRF measurements given by a Niton XL-5 device to minimize deviations at low energies. The final model was benchmarked to an experimental XRF measurement of Cr-coated Zr4 taken by the XL-5. The percent error in the resulting XRF peak intensities was within <span><math><mrow><mo>±</mo></mrow></math></span> 4.92% for the K<sub>α1</sub> peaks and within <span><math><mrow><mo>±</mo></mrow></math></span> 16.0% for the K<sub>β1</sub>. The discrepancies in the magnitude of these errors are largely due to the K<sub>β1</sub> peaks having far fewer counts in the spectra that were compared. Nonetheless, these results demonstrate the potential for MCNP 6.2 to accurately predict low-energy X-ray interactions such as XRF. The deviations observed were similar to those seen in the 0.1–1 MeV range in prior works, despite only being in the <span><math><mrow><mo>∼</mo></mrow></math></span> 5–30 keV range themselves.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105487"},"PeriodicalIF":3.3,"publicationDate":"2024-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142434141","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Large-eddy simulation of turbulent flow and heat transfer of helically corrugated tubes in the intermediate heat exchanger of a very-high-temperature gas-cooled reactor 超高温气冷式反应堆中间热交换器中螺旋波纹管的湍流和传热的大涡流模拟

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-14 DOI: 10.1016/j.pnucene.2024.105488

Qingxiang Hu, Haifeng Liu, Qi Sun, Xiaozhong Wang, Jie Wang, Wei Peng

{"title":"Large-eddy simulation of turbulent flow and heat transfer of helically corrugated tubes in the intermediate heat exchanger of a very-high-temperature gas-cooled reactor","authors":"Qingxiang Hu, Haifeng Liu, Qi Sun, Xiaozhong Wang, Jie Wang, Wei Peng","doi":"10.1016/j.pnucene.2024.105488","DOIUrl":"10.1016/j.pnucene.2024.105488","url":null,"abstract":"<div><div>The intermediate heat exchanger (IHX) is a vital component of very-high-temperature gas-cooled reactors (VHTRs) utilized for thermal applications of nuclear energy, specifically for hydrogen production. Enhancing the heat transmission capacity of IHXs is essential to provide sufficient heat for thermal processes. This study uses large eddy simulations to investigate IHX models consisting of both a smooth circular tube and helically corrugated tubes with five different geometric parametrizations. The results show how turbulent flow and heat transfer depend on the geometric parameters. Based on theories such as boundary layer theory, field synergy, and extreme dissipation, the characteristics of boundary layer separation, secondary flow, turbulent transport, field synergy, and dissipation characteristics in helically corrugated tubes are quantitatively analyzed. The study also investigates enhanced heat transfer mechanisms within the helically corrugated tubes, and the results attribute the enhanced heat transfer in helically corrugated tubes to the helical structure, which hinders the development of a fluid boundary layer, strengthens the intensity of secondary flow and turbulent transport, improves the synergy between the fluid velocity field and the temperature gradient, and reduces the thermal potential energy loss during the fluid heat transfer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105488"},"PeriodicalIF":3.3,"publicationDate":"2024-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142434140","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Thermal-hydraulic analysis of helical coil once-through steam generators under complex oceanic conditions 复杂海洋条件下螺旋线圈直通式蒸汽发生器的热工水力分析

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105484

Chao Zhang, Jiangping Li, Zihan Liu, Yandong Hou, Yunlong Zhou

{"title":"Thermal-hydraulic analysis of helical coil once-through steam generators under complex oceanic conditions","authors":"Chao Zhang, Jiangping Li, Zihan Liu, Yandong Hou, Yunlong Zhou","doi":"10.1016/j.pnucene.2024.105484","DOIUrl":"10.1016/j.pnucene.2024.105484","url":null,"abstract":"<div><div>The helical coil once-through steam generator (HCOTSG) possesses several advantages, including excellent thermal efficiency, compact size, rapid start-up, and low energy consumption. These qualities make it particularly well-suited for use in ship reactors. This paper introduces a one-dimensional thermal-hydraulic calculation program for helical coil once-through steam generators (HCOTSGs) operating under complex oceanic conditions, utilizing the finite difference method alongside fixed boundary conditions and staggered grid techniques. By incorporating the heat transfer and friction relations to the helical coil, as well as considering the geometric structure of the helical coil when incorporating ocean conditions, the characteristics of the helical coil are reflected in the one-dimensional code. The impact of oceanic conditions is considered by modifying the momentum equation. This paper considers highly complex oceanic conditions that involve a coupling of inclining and oscillation. Furthermore, the accuracy of the developed program is verified by comparing the HCOTSG design parameters of the MRX reactor with the calculated values from internationally recognized prediction programs against the findings presented here.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423748","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Oxygen embrittlement Zircaloy-4 燃料销在模拟失壳事故条件下失效：氧脆性

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105485

Saurabh Sagar , Mohd Kaleem Khan , Manabendra Pathak , Suparna Banerjee , Tapan Kumar Sawarn , S.K. Yadav , R.N. Singh

{"title":"Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Oxygen embrittlement","authors":"Saurabh Sagar , Mohd Kaleem Khan , Manabendra Pathak , Suparna Banerjee , Tapan Kumar Sawarn , S.K. Yadav , R.N. Singh","doi":"10.1016/j.pnucene.2024.105485","DOIUrl":"10.1016/j.pnucene.2024.105485","url":null,"abstract":"<div><div>An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423749","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Derivative dynamic time warping algorithm with introduced correction for varying load fault diagnosis of nuclear power system steam turbine units 用于核电系统蒸汽轮机组不同负荷故障诊断的导入校正的衍生动态时间扭曲算法

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105490

Haotong Wang, Yanjun Li, Guolong Li, Shengdi Sun, Baozhi Sun, Yuanwei Cao, Jianxin Shi

{"title":"Derivative dynamic time warping algorithm with introduced correction for varying load fault diagnosis of nuclear power system steam turbine units","authors":"Haotong Wang, Yanjun Li, Guolong Li, Shengdi Sun, Baozhi Sun, Yuanwei Cao, Jianxin Shi","doi":"10.1016/j.pnucene.2024.105490","DOIUrl":"10.1016/j.pnucene.2024.105490","url":null,"abstract":"<div><div>The Derivative Dynamic Time Warping (DDTW) algorithm is improved to address the multi-parameters time series classification problem faced by nuclear power system steam turbine units varying load fault diagnosis. Firstly, the entire load changing process is treated as a single sample rather than multiple time-step-samples. This ensures the complete information on the load changing processes, while avoiding interference from normal data fluctuations during faults. Secondly, Time Series Position Coefficient and Time Series Length Coefficient are proposed to correct the DDTW algorithm from two perspectives: the sequences lengths and the data positions in the sequences. This solves the singularities and timeline scaling problems, thereby preventing interference introduced by data sequences' lengths differences and ''similar data appearing at different times'' problem. The nuclear power system steam turbine unit simulation model was built to obtain load changing processes data under normal and faults statuses. In the varying load fault diagnosis test based on these data, the improved DDTW algorithm achieved an accuracy of 1.38%–12.06% higher than other methods, reaching 87.50%. Finally, The Deep Convolutional Generative Adversarial Networks (DCGAN) model was used to generate data to supplement the limited samples of complete load changing processes, and the accuracy of the novel method increased to 95.51% with the increase of data used to support the comparison.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":"Article 105490"},"PeriodicalIF":3.3,"publicationDate":"2024-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142433222","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Interaction behaviour of alloy 690 upon exposure to P2O5 containing borosilicate glass at simulated vitrification conditions 合金 690 在模拟玻璃化条件下暴露于含 P2O5 的硼硅玻璃时的相互作用行为

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-11 DOI: 10.1016/j.pnucene.2024.105480

Mohit Rattanpal , Poulami Chakraborty , Kulwant Singh , Soudamini N , R.K. Mishra , Pranesh Sengupta , R. Tewari

{"title":"Interaction behaviour of alloy 690 upon exposure to P2O5 containing borosilicate glass at simulated vitrification conditions","authors":"Mohit Rattanpal , Poulami Chakraborty , Kulwant Singh , Soudamini N , R.K. Mishra , Pranesh Sengupta , R. Tewari","doi":"10.1016/j.pnucene.2024.105480","DOIUrl":"10.1016/j.pnucene.2024.105480","url":null,"abstract":"<div><div>This study investigates the interaction between alloy 690 and two types of glasses: pristine borosilicate and P<sub>2</sub>O<sub>5</sub>-containing borosilicate glass, at typical glass pouring temperatures encountered during the vitrification of high-level radioactive waste in the back-end of Nuclear Fuel Cycle (NFC). Partial crystallization of both glasses was observed at the alloy 690/glass interface, with certain, though not entirely identical, crystalline phases forming at the interface. The density of these crystalline phases is significantly higher than that of the surrounding glass, which raises concerns about these reaction products settling at the bottom of the furnace. Such sedimentation could potentially obstruct the freeze valve, thereby halting the vitrification process. Additionally, intergranular grooves on the alloy surface exposed to P<sub>2</sub>O<sub>5</sub>-containing borosilicate glass were found to disappear with prolonged exposure. This phenomenon is attributed to the strong corrosive action of the highly basic and oxidizing P<sub>2</sub>O<sub>5</sub>-bearing glass, leading to the peeling away of the entire exposed surface of alloy 690.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":"Article 105480"},"PeriodicalIF":3.3,"publicationDate":"2024-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423751","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study of the effect of seawater properties on the performance of molten fuel fragmentation 研究海水特性对熔融燃料破碎性能的影响

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-10 DOI: 10.1016/j.pnucene.2024.105483

Yinmeng Zhang , Shaojie Tan , Yangkai Huang , Songbai Cheng , Hui Cheng

{"title":"Study of the effect of seawater properties on the performance of molten fuel fragmentation","authors":"Yinmeng Zhang , Shaojie Tan , Yangkai Huang , Songbai Cheng , Hui Cheng","doi":"10.1016/j.pnucene.2024.105483","DOIUrl":"10.1016/j.pnucene.2024.105483","url":null,"abstract":"<div><div>In this study, in order to explore the effects of seawater properties on the fragmentation behavior of melt jets during a severe core meltdown accident in the light water reactor, visualized fragmentation experiments are carried out by releasing superheated melt into subcooled water at different coolant salinities, melt temperatures, water temperatures, and melt penetration velocities using the VTMCI (Visualized Thermo-hydraulic characteristics in Melt Coolant Interaction) facility at Sun Yat-Sen University. It is found that under the current experimental conditions, as the coolant salinity increases, the size of the debris decreases, while the variation of debris sphericity and debris bed porosity is insignificant. When the water temperature or melt temperature increases, the sphericity of the debris is higher, and the porosity of the debris bed and the size of the debris decrease. When the penetration rate of the melt is higher, smaller particles can be generated, but it has no significant impact on the debris bed porosity and debris sphericity. The Weber number theory can be used to predict the median diameter of debris, while the Stephan number (<em>St</em>) can be used to predict the trends in debris bed porosity and debris sphericity. This study contributes to a deeper understanding of the actual fragmentation process mechanism of molten materials in severe accidents of light water reactors. The experimental data obtained will also contribute to the development, validation, and improvement of relevant physical models in China's pressurized water reactor severe accident analysis codes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":"Article 105483"},"PeriodicalIF":3.3,"publicationDate":"2024-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423750","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Linked calculation method based on next event estimation for directional shielding problems 基于下一事件估计的定向屏蔽问题关联计算方法

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-08 DOI: 10.1016/j.pnucene.2024.105465

Huayang Zhang , Bin Zhong , Ruoyang Qi , Longfei Xu , Huayun Shen , Jinhong Li

引用次数: 0

Sensitivity and uncertainty analysis of neutron activation source terms based on statistical sampling method 基于统计采样法的中子活化源项敏感性和不确定性分析

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2024-10-08 DOI: 10.1016/j.pnucene.2024.105472

Xiao Yu, Jingyu Zhang, Yixin Xiao, Wangmuhong Ni, Juanjuan Wang

{"title":"Sensitivity and uncertainty analysis of neutron activation source terms based on statistical sampling method","authors":"Xiao Yu, Jingyu Zhang, Yixin Xiao, Wangmuhong Ni, Juanjuan Wang","doi":"10.1016/j.pnucene.2024.105472","DOIUrl":"10.1016/j.pnucene.2024.105472","url":null,"abstract":"<div><div>The calculation of neutron activation source terms is an essential task in the radiation of nuclear reactors, and its sensitivity and uncertainty (S&U) analysis can help better identify and control the influencing factors and safety margin of activation source terms. In this paper, we studied the statistical sampling strategies for the S&U analysis of activation source terms, established the calculation process of sensitivity coefficient and uncertainty, and the S&U analysis module based on sampling methods was integrated for the proprietary developed activation computation program ABURN. Two sampling ways, Monte Carlo and Latin hypercube sampling treatments, were employed to figure out the cases of pure decay single chain, pure activation single chain, and complex reaction network. The sensitivity coefficients and uncertainty of nuclide inventory were evaluated by taking the uncertainty of decay constant or reaction cross-section into account. The simulation results indicate that after introducing the LHS sampling method, the activation calculation program ABURN can improve computational efficiency in sensitivity and uncertainty analysis. Subsequent results demonstrated that the sensitivity coefficients and uncertainty solutions based on the sampling methods in the ABURN program were in good agreement with the analytical solutions and the FISPACT program, which verified the accuracy of the approaches and program in this paper.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":"Article 105472"},"PeriodicalIF":3.3,"publicationDate":"2024-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423746","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0