Nuclear Engineering and Design最新文献_第2页

Qusai-Static autonomous load following analysis of a small modular lead-bismuth cooled fast reactor 一个小型模块化铅铋冷却快堆的静态自主负荷分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-14 DOI: 10.1016/j.nucengdes.2025.114122

Kefan Zhang, Wenshun Duan, Weixiang Wang, Sifan Dong, Hongli Chen

引用次数: 0

Techno-economic assessment of a smTMSR-based nuclear–renewable hybrid energy system with high-temperature steam electrolysis 基于smtmsr的高温蒸汽电解核可再生混合能源系统的技术经济评价

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-14 DOI: 10.1016/j.nucengdes.2025.114138

Xueying Nie , Maosong Cheng , Xiandi Zuo , Zhimin Dai

{"title":"Techno-economic assessment of a smTMSR-based nuclear–renewable hybrid energy system with high-temperature steam electrolysis","authors":"Xueying Nie , Maosong Cheng , Xiandi Zuo , Zhimin Dai","doi":"10.1016/j.nucengdes.2025.114138","DOIUrl":"10.1016/j.nucengdes.2025.114138","url":null,"abstract":"<div><div>The fluctuation nature of renewable energy causes the imbalance between energy demand and supply. The nuclear-renewable hybrid energy system (NRHES) that couples renewable energy with a small modular thorium molten salt reactor (smTMSR), thermal energy storage (TES), and high-temperature steam electrolysis (HTSE) is constructed to solve this problem. Three operation modes (Mode 1: NRHES without HTSE, Mode 2: NRHES coupled with HTES, Mode 3: NRHES coupled with HTES with limiting condition) with different energy management strategies are proposed and compared firstly in current research. The multi-objective particle swarm optimization (MOPSO) with better performance is chosen to conduct multi-objective capacity configuration optimization of the NRHES. The optimization objectives include minimizing deficiency of power supply probability (DPSP), energy waste possibility (EWP), and levelized cost of energy (LCOE). The three-dimensional Pareto solutions are obtained and the mode 3 is demonstrated to be the optimal mode. The results analysis shows that the addition of HTSE with limiting condition could increase proportions of power supply and hydrogen production of nuclear energy, which reduces the wasted energy naturally. This study provides valuable reference information for relieving the imbalance of energy supply and demand in NRHES effectively.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114138"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143948572","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

Seismic soil-structure – interaction analysis of a radioactive waste repository considering on site measurements and subsoil improvement 考虑现场测量和底土改良的放射性废物处置库地震土-结构相互作用分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-14 DOI: 10.1016/j.nucengdes.2025.114137

Sotiria Karapetrou , Marco Hildmann , Fritz-Otto Henkel

{"title":"Seismic soil-structure – interaction analysis of a radioactive waste repository considering on site measurements and subsoil improvement","authors":"Sotiria Karapetrou , Marco Hildmann , Fritz-Otto Henkel","doi":"10.1016/j.nucengdes.2025.114137","DOIUrl":"10.1016/j.nucengdes.2025.114137","url":null,"abstract":"<div><div>Aim of this paper is to present the modelling and analysis steps for the evaluation and design of a repository structure, which will be used as a long-term storage for low and intermediate radioactive waste. Its initial design began in the early 90ies but it was not concluded at that time. As the seismic requirements increased significantly over the past years, the new design of the construction is required. Two analysis cases are implemented considering both the soil – structure interaction effects. In the first case, the soil –structure system is considered in one single model implementing the substructuring method. In the second case the soil – structure interaction effects are determined in a two-step modelling procedure by generating at first appropriate soil springs and introducing them continuously to the numerical model of the structure. These two separate analysis procedures are required in order to achieve an integrated evaluation and design process not only of the repository structure itself but also for the various components.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114137"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143948571","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

Investigation into a Broyden-Type method for the Fully-Implicit scheme of Two-Fluid model and its convergence performance 双流体模型全隐式格式的broyden型方法及其收敛性能研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-14 DOI: 10.1016/j.nucengdes.2025.114141

Hao Zhang, Meng Zhao, Yanhua Yang

{"title":"Investigation into a Broyden-Type method for the Fully-Implicit scheme of Two-Fluid model and its convergence performance","authors":"Hao Zhang, Meng Zhao, Yanhua Yang","doi":"10.1016/j.nucengdes.2025.114141","DOIUrl":"10.1016/j.nucengdes.2025.114141","url":null,"abstract":"<div><div>The fully-implicit scheme of two-fluid model enables calculation with large time step, and it is attractive for the long problem time simulation and calculation with fine cells. One of the key aspects of the solution algorithm is the calculation of Jacobian matrix. First, the Jacobian matrix of fully-implicit scheme is ill-conditioned. We find that non-dimensionalizing the governing equations and primary variables can reduce the condition number effectively. Second, computing the Jacobian matrix of the fully-implicit scheme is time-consuming. To address this issue, we adopt Broyden-Schubert method. This method is not only easily implementable and has a fast computational speed but also can maintain the sparse structure of the matrix. However, it leads to smaller convergence region and lower convergence rate. A very natural idea is to calculate Jacobian matrix using direct calculation method for the first few steps, and then employ Broyden-Schubert method for the remaining steps. It is found that a small number of direct calculation iterations can significantly improve convergence performance. Therefore, this hybrid method may be a potential development direction for the fully-implicit scheme. It is important to note that the conclusions presented in this paper are derived from near-steady-state and relatively simple transient cases. As a result, the applicability of the Broyden-type method to complex two-phase flow transient cases cannot be guaranteed. Therefore, further test of its applicability in two-phase flow is essential and needs to be conducted in future research work.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114141"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143947811","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

Effect of rolling motions on the single and two-phase pressure drop in annulus channel with helical-finned heater for floating nuclear reactor 滚动运动对浮动核反应堆螺旋翅片加热器环空通道单相和两相压降的影响

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-14 DOI: 10.1016/j.nucengdes.2025.114092

Hyukjae Ko , Heepyo Hong , Ja Hyun Ku , Goon-Cherl Park , Hyoung Kyu Cho

{"title":"Effect of rolling motions on the single and two-phase pressure drop in annulus channel with helical-finned heater for floating nuclear reactor","authors":"Hyukjae Ko , Heepyo Hong , Ja Hyun Ku , Goon-Cherl Park , Hyoung Kyu Cho","doi":"10.1016/j.nucengdes.2025.114092","DOIUrl":"10.1016/j.nucengdes.2025.114092","url":null,"abstract":"<div><div>Interest in utilizing nuclear power in the maritime environment has been growing. However, the limited availability of research on floating nuclear power plants poses challenges to support the design and safety evaluation process. To address this gap, experimental platforms were developed in Seoul National University to conduct thermal–hydraulic experiments under simulated rolling conditions. To replicate pressurized water reactor conditions, a compact experimental loop using refrigerant R134a as a simulant fluid was designed and mounted on the rolling platform. An annulus geometry test-section was utilized, to imitate the rod-centered subchannel of the reactor core. In this study helical-finned heater rod was used representing one of the potential fuel rod geometries for tight lattice cores. During the experiments, motion-driven flow fluctuations were constrained within ±1 % range to neglect the flow fluctuation effect on pressure drop. The experimental results confirmed that pressure drop is affected by motion conditions. Under inclined conditions, the pressure drop decreased, primarily due to the reduction in elevation between the test-section inlet and outlet. On the other hand, under rolling conditions, the pressure drop fluctuated, driven mainly by oscillations in acceleration, including both gravitational and centrifugal components. Furthermore, the range of pressure drop fluctuations was found to exceed what could be attributed solely to acceleration fluctuations. This additional increase was estimated to result from frictional pressure drop fluctuation occurring under the rolling conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114092"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143948574","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

Investigation of low-frequency guided waves in a fluid-filled pipe submerged in LBE LBE中充液管道低频导波的研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-13 DOI: 10.1016/j.nucengdes.2025.114118

Yu Jiang , Genshan Jiang , Yu Zhou , Hao Li

{"title":"Investigation of low-frequency guided waves in a fluid-filled pipe submerged in LBE","authors":"Yu Jiang , Genshan Jiang , Yu Zhou , Hao Li","doi":"10.1016/j.nucengdes.2025.114118","DOIUrl":"10.1016/j.nucengdes.2025.114118","url":null,"abstract":"<div><div>This paper investigates the propagation characteristics of acoustic waves in liquid-filled pipes submerged in lead–bismuth eutectic(LBE). Expressions for the phase velocities of the fluid-dominated (s = 1) and shell-compressed (s = 2) waves in a fluid-filled pipe submerged in LBE are derived based on the Kennard equation, then, the effects of pipe radius, wall thickness, and temperature on the wave speed and attenuation of low-frequency such waves are analyzed via numerical simulations. The results show that LBE has less impact on the s = 1 phase velocity and a pronounced influence on the s = 2 phase velocity and the attenuation of the two waves. The s = 1 phase velocity is correlated positively with the thickness-to-radius ratio, and the s = 2 phase velocity is correlated negatively with the thickness-to-radius ratio. As the thickness-to-radius ratio increases, the acoustic attenuation in the fluid-filled pipe decreases. Increasing the temperature decreases the phase velocity of both the s = 1 and s = 2 waves. Comparing the pipe-wall radial vibration displacements induced by the two waves shows that the pipe-wall displacement caused by the fluid-dominated wave is greater, and it serves as the primary carrier of the acoustic signal during leakage. The present results contribute to the application and development of acoustic inspection techniques for the health monitoring of heat-exchange pipes in the steam generators of lead–bismuth fast reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114118"},"PeriodicalIF":1.9,"publicationDate":"2025-05-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935295","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

Thermo-elastoplastic modeling of UO2/Zr dispersion fuel UO2/Zr分散燃料的热弹塑性模型

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-13 DOI: 10.1016/j.nucengdes.2025.114129

S. Khajian, A. Zolfaghari, Z. Kowsar

{"title":"Thermo-elastoplastic modeling of UO2/Zr dispersion fuel","authors":"S. Khajian, A. Zolfaghari, Z. Kowsar","doi":"10.1016/j.nucengdes.2025.114129","DOIUrl":"10.1016/j.nucengdes.2025.114129","url":null,"abstract":"<div><div>Dispersed fuel elements are widely utilized in research and test reactors. The high thermal conductivity of dispersion fuels allows a significant reduction of temperature in comparison to a standard oxide fuel. The main objective of this paper is a simulation of the displacement and the associated stresses using the governing equations for mechanical equilibrium, heat generation, heat transfer, stress–strain relationship, swelling, thermal expansion, fission-induced creep, created pressure by fission gas bubbles in a plate type cermet fuel, <span><math><msub><mrow><mi>UO</mi></mrow><mn>2</mn></msub></math></span> ceramic fuel in <span><math><mrow><mi>Zr</mi></mrow></math></span> as a metallic matrix, which is subjected to thermal and mechanical loadings. The mathematical derivations are obtained for an equivalent unit sphere cell containing an inner layer for fuel particles and an outer layer for matrix. The outcome stresses are verified against reference solutions. To implement the algorithm, a code-named Kowsar-2 is developed in which a fuel plate is discretized in the axial and through-thickness direction in the first step. Then the thermal calculation is carried out for each segment considering the effect of solid fission products, fission gases released over time along fission gas diffusion, bubble production, swelling, etc. Temperature of coolant, oxide layer, clad, and fuel are also obtained in this step. Mechanical treatments are followed for a typical cell to get stress and strain distributions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114129"},"PeriodicalIF":1.9,"publicationDate":"2025-05-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935296","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

LFR related R&D in the Dutch PIONEER program 荷兰先锋项目LFR相关研发

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-12 DOI: 10.1016/j.nucengdes.2025.114146

Ferry Roelofs, Paul Breijder, Victor Habiyaremye, Mohammed Muaaz Mohamed Dilshad Hussain, Akshat Mathur, Fitriana Nindiyasari, Fajar Pangukir, Marek Stempniewicz, Sander van Til, Heleen Uitslag-Doolaard, Dirk Visser, Eva de Visser-Týnová, Kevin Zwijsen

{"title":"LFR related R&D in the Dutch PIONEER program","authors":"Ferry Roelofs, Paul Breijder, Victor Habiyaremye, Mohammed Muaaz Mohamed Dilshad Hussain, Akshat Mathur, Fitriana Nindiyasari, Fajar Pangukir, Marek Stempniewicz, Sander van Til, Heleen Uitslag-Doolaard, Dirk Visser, Eva de Visser-Týnová, Kevin Zwijsen","doi":"10.1016/j.nucengdes.2025.114146","DOIUrl":"10.1016/j.nucengdes.2025.114146","url":null,"abstract":"<div><div>The Netherlands is preparing for an expansion of its nuclear energy generation capacity with two or even four large light water reactors and possible deployment of small modular reactors. In order to ensure nuclear energy generation capacity not only on the short term (the 21st century) but far beyond that, closing the fuel cycle and making much more efficient use of the earth’s natural resources is a requirement. There are different routes towards closing the fuel cycle. In this paper, the R&D activities on liquid metal fast reactors and in particular lead fast reactors, performed within the Dutch PIONEER research program, are highlighted.</div><div>The design and safety analysis activities mainly deal with thermal hydraulics topics, ranging from system thermal hydraulics code applications to component level CFD simulations. With respect to fuel and material activities, the unique experimental facilities of NRG PALLAS, are being used for R&D activities that can’t easily be carried out at other research facilities. Unique data from various irradiation campaigns are under analysis and dissolution and reprocessing studies for fast reactor fuel are being carried out. Most of the work is embedded in European collaborative framework projects, thus contributing to the European developments on LFR reactor and related fuel cycle technology.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114146"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935293","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

Outcome of the VIKING project: status and perspectives of numerical modeling of flow-induced vibrations of nuclear power plant components VIKING项目成果：核电厂部件流致振动数值模拟的现状与展望

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-12 DOI: 10.1016/j.nucengdes.2025.114131

K. Zwijsen , S. Tajfirooz , F. Roelofs , A. Papukchiev , D. Vivaldi , H. Hadžić , S. Benhamadouche , W. Benguigui , T. Norddine , H. Iacovides , A. Cioncolini , M.R.A. Nabawy , K. Angele , E. Lillberg , B. Chazot , E. Iyamabo

{"title":"Outcome of the VIKING project: status and perspectives of numerical modeling of flow-induced vibrations of nuclear power plant components","authors":"K. Zwijsen , S. Tajfirooz , F. Roelofs , A. Papukchiev , D. Vivaldi , H. Hadžić , S. Benhamadouche , W. Benguigui , T. Norddine , H. Iacovides , A. Cioncolini , M.R.A. Nabawy , K. Angele , E. Lillberg , B. Chazot , E. Iyamabo","doi":"10.1016/j.nucengdes.2025.114131","DOIUrl":"10.1016/j.nucengdes.2025.114131","url":null,"abstract":"<div><div>Crucial nuclear power plant (NPP) components, such as fuel assemblies and steam generators, are exposed to flow-induced vibrations (FIV), potentially leading to fatigue problems and fretting wear of the material. Damage or failure of these components may lead to safety issues, thereby potentially necessitating unplanned outages of the reactor, resulting in substantial repair and standstill costs. With FIV being one of the leading causes of damage to these components, it is important to assess its impact on the integrity of fuel rods and steam generator tubes during the early design phase. While such an assessment has historically been done using semi-empirical models, due to the rise in computing power and capabilities, numerical tools are used more frequently, in particular in the last 10–15 years. To assess and further advance the current state-of-the-art of studying FIV in NPPs, the joint industry VIKING (Vibration ImpaKt In Nuclear power Generation) project was launched at the beginning of 2020. In this project, nine organizations collaborated for almost four years on FIV of configurations representative of steam generators and fuel rods and assemblies. This was done by performing numerical benchmark studies on five different experimental facilities. The current paper describes the main results and conclusions obtained from each numerical benchmark. Based on the individual findings, the status and perspectives of numerically simulating FIV of the aforementioned NPP components are presented.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114131"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935297","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

Effect of tuned mass dampers on seismic fragility of piping system in nuclear power plant – Integrating experiment and finite element analysis 调谐质量阻尼器对核电站管道系统地震易损性的影响——试验与有限元分析相结合

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-05-12 DOI: 10.1016/j.nucengdes.2025.114133

Bu-Seog Ju , Shinyoung Kwag , Sangwoo Lee

{"title":"Effect of tuned mass dampers on seismic fragility of piping system in nuclear power plant – Integrating experiment and finite element analysis","authors":"Bu-Seog Ju , Shinyoung Kwag , Sangwoo Lee","doi":"10.1016/j.nucengdes.2025.114133","DOIUrl":"10.1016/j.nucengdes.2025.114133","url":null,"abstract":"<div><div>Nuclear power plants encompass complex piping systems, which play a critical role in both operation and maintenance. With the occurrence of various beyond design earthquakes worldwide, there is an increasing necessity to enhance the seismic safety of the key components. Tuned Mass Dampers (TMDs) have gained prominence as an effective strategy for improving the seismic performance of piping systems. Existing studies on TMDs have primarily focused on evaluating their performance based on the reduction of system responses, such as displacement and acceleration, at locations where these responses are maximized. However, in actual nuclear power plant structures, the seismic evaluation of piping system is typically assessed through fragility analysis, and most piping systems are more susceptible to localized damage at connection points, such as elbows and T-joints, rather than at points of the maximum response. This study investigates the effectiveness of TMDs under various seismic loadings and assesses their impact on the seismic fragility curves of piping systems. To achieve this, an extended numerical experiment is conducted, developing a numerical model of a full-scale piping system that was validated against experimental results. Additionally, we examined the ability of TMDs to mitigate various local piping responses under high-intensity earthquakes which are challenging to address experimentally. Finally, fragility analyses are performed using several previously proposed TMD design methods, evaluating the influence of TMDs on the seismic fragility curves of the piping system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114133"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935294","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