Nuclear Engineering and Design最新文献

{"title":"A preferable siliconized graphite with high graphite content in thrust bearings of shaft-sealed main coolant pump","authors":"S.H. Liu ,&nbsp;T.H. Liang ,&nbsp;B.J. Zhang ,&nbsp;J.Q. An ,&nbsp;L. Cai ,&nbsp;M.K. Lei","doi":"10.1016/j.nucengdes.2026.114831","DOIUrl":"10.1016/j.nucengdes.2026.114831","url":null,"abstract":"<div><div>A challenge of achieving high load capacity and low wear loss for siliconized graphite counterparts of the thrust bearings in shaft-sealed nuclear main pump of pressurized water reactor power plant is urgent due to the mixed lubrication under high contact pressure × sliding velocity (<em>pv</em>) factor above 65.0 MPa·m/s in high-temperature and high-pressure water. Wear and lubrication of siliconized graphite 42C-44SiC-14Si (vol%) with elevated graphite content were investigated using self-mated pairs in the high-temperature and high-pressure tribometer on a pin-on-disc configuration. Contact pressure <em>p</em> of 0.87 and 2.08 MPa with sliding velocity <em>v</em> from 0.5 to 33.1 m/s supply the <em>pv</em> factors of 0.4–68.8 MPa·m/s in 50 °C and 1.5 MPa water. A tribology-induced oxidation mechanism on SiC of siliconized graphite is proposed under the mixed lubrication. The selective oxidation of SiC characterizes the supersaturation of oxygen, fragmentation of SiC, and amorphization of silicon oxide. Oxidized SiC peaks alternating with valleys of non-oxidized graphite and Si phases facilitate the regular and corrugated-like texture on mated surfaces that promotes hydrodynamic effect. The lower wear depth of 0.12 μm and coefficient of friction (COF) of 0.015 are achieved under 68.8 MPa·m/s. Siliconized graphite 42C-44SiC-14Si with high graphite content provides the solid lubricant under boundary lubrication, and the uniform pressure distribution of water film under mixed lubrication. Therefore, the siliconized graphite as a promising material enables a long service life of thrust bearings in demanding nuclear applications, withstanding high load under operating conditions and offering superior resistance to more startup-shutdown cycles under specific conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114831"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190249","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}

{"title":"Modelling of MSRE graphite temperature in porous-medium multi-physics simulations","authors":"S. Amirkhosravi ,&nbsp;A. Scolaro ,&nbsp;F. van Niekerk ,&nbsp;M.H. du Toit ,&nbsp;A. Pautz","doi":"10.1016/j.nucengdes.2026.114757","DOIUrl":"10.1016/j.nucengdes.2026.114757","url":null,"abstract":"<div><div>MSRs stand out as prominent candidates among advanced reactor designs, addressing the global demand for safer and more sustainable nuclear energy. Accurate multi-physics modelling is essential for the advancement of MSR technology, particularly for understanding the thermos-hydraulic behaviour of graphite under irradiation. This study focuses on developing and implementing a high-fidelity methodology within the GeN-Foam code to model graphite temperature distribution within porous-medium multi-physics simulations, using the MSRE as a benchmark. The approach combines thermal-hydraulic and neutronic modelling by using Serpent-generated cross-section data as inputs for the <strong>Gen-Foam</strong> neutronic solver. Validation against MSRE measurements performed at ORNL benchmark data confirms the framework's reliability. The axial temperature distribution yields a Mean Absolute Percentage Error (MAPE) of 1.09%, while the radial distribution shows a MAPE of 0.62%. The average graphite temperature of 935.6 K is consistent with the ORNL reference value of 936.4 K under steady-state conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114757"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078927","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}

{"title":"Research on fluid-structure interaction characteristics of transient pressure waves in reactor coolant pump under shaft seizure accident","authors":"Teng Niu ,&nbsp;Yi Bin Li ,&nbsp;Hai Long Yuan ,&nbsp;Xue Zhao ,&nbsp;Kong Sheng Liu","doi":"10.1016/j.nucengdes.2026.114795","DOIUrl":"10.1016/j.nucengdes.2026.114795","url":null,"abstract":"<div><div>This study investigates the fluid-structure interaction (FSI) characteristics of transient pressure waves during a reactor coolant pump (RCP) shaft seizure accident (SSA) through bidirectional FSI numerical simulation of the coolant pipeline. Based on a model of the HPR1000 reactor single-loop system with matched resistance characteristics, the analysis focuses on the RCP flow field pressure, internal pressure fluctuations, and pipeline dynamic response. The results demonstrate that coolant pipeline fluid-structure interaction (CPFSI) significantly alters pressure distributions in RCP flow components. During shaft seizure, CPFSI causes a notable expansion of the low-pressure zone at the impeller inlet and an increase in volute pressure. Immediately after shaft seizure, it induces a widespread pressure decrease at the inlet nozzle, a significant enlargement of the low-pressure region within the guide vane flow passage, and a slight expansion of the low-pressure area at the volute outlet. After shaft seizure ends, CPFSI leads to substantial pressure reductions at the inlet nozzle, impeller inlet, volute annular cavity, and volute outlet, alongside a marked expansion of low-pressure zones at the impeller inlet and IPS and a contraction of the high-pressure zone at the GPS inlet. Throughout the shaft seizure transition process, the transition pipe experiences the most pronounced deformation and fluctuation, followed by the hot leg pipe, with the cold leg pipe showing minimal variation. These structural vibrations intensify pressure fluctuations at RCP monitoring points, leading to either attenuation or amplification of transient pressure wave amplitudes. This research reveals the coupling mechanism between transient pressure waves and pipeline dynamics during an SSA, providing a theoretical basis for accurately assessing RCP operational safety.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114795"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081263","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}

{"title":"Study on the base isolation design and parameter optimization analysis of friction pendulum bearings for reactor building in swimming pool-type low-temperature heating reactor","authors":"Yingying Gan ,&nbsp;Xiaoying Sun ,&nbsp;Pengxiang Dong ,&nbsp;Ziqiao Liu","doi":"10.1016/j.nucengdes.2025.114739","DOIUrl":"10.1016/j.nucengdes.2025.114739","url":null,"abstract":"<div><div>The swimming pool-type low-temperature heating reactor (SPLTHR) is a single-unit small heating reactor that can serve as an alternative to fossil energy. The peak ground acceleration (PGA) for the safe shut- down earthquake (SSE) at the proposed site is up to 0.5 g in horizontal direction. To ensure seismic safety and improve economic efficiency of the reactor, the Friction Pendulum (FP) bearing is employed for the base isolation design of the reactor building. Firstly, a three-dimensional finite element model (FEM) of the reactor building is established. The layout scheme of the base isolation layer is designed. Subsequently, a parameter optimization analysis about the equivalent radius of curvature and dynamic friction coefficient of the FP bearing is conducted to achieve the optimal isolation performance for the reactor building. Finally, the acceleration response spectrum (ARS) in three directions were compared between the base-isolated system and non- isolated system at the same place. The acceleration reduction rate was defined to quantified the isolation performance. The study results indicate that the base isolation layer using 28 FP bearings with load capacity in axial direction of 15,000 kN and 8 viscous damper can meet the design requirement. The dynamic friction coefficient of the FP bearing has a more significant influence on the isolation performance than the equivalent radius of curvature. In general, a larger equivalent radius of curvature and a smaller dynamic friction coefficient result in better isolation performance. The ARS in horizontal direction of the superstructure in the non-isolated system completely envelops that of the base-isolated system. The seismic response of the base-isolated system shows a substantial reduction in the dominant frequency and a significant decrease in the ARS of the superstructure in horizontal direction. The maximum reduction rates for zero-period acceleration (ZPA) and peak acceleration can reach up to 75.0 % and 85.4 %, respectively, demonstrating excellent isolation performance. Compared to the ARS in vertical direction of the non-isolated system, the base-isolated system has a lower dominant frequency, a leftward shift in peak acceleration response (with lower peak frequency), and an insignificant increase in peak values. It is recommended to focus on the seismic response of key equipment which is sensitive to the vertical frequency ranges of the base-isolated system, and to implement appropriate local vertical isolation measures if necessary.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114739"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081255","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}

{"title":"Modeling considerations for passive safety systems in Korean SMRs using system analysis codes","authors":"Seong-Su Jeon,&nbsp;Jungjin Bang,&nbsp;Sang Gyun Nam,&nbsp;Jehee Lee,&nbsp;Youngjae Park,&nbsp;Soon-Joon Hong","doi":"10.1016/j.nucengdes.2025.114747","DOIUrl":"10.1016/j.nucengdes.2025.114747","url":null,"abstract":"<div><div>Numerous Small Modular Reactors (SMRs) are being developed worldwide, and they are equipped with various types of Passive Safety Systems (PSSs). In the Republic of Korea, SMART100 and i-SMR are representative SMRs. SMART100 includes the Passive Safety Injection System (PSIS), the Passive Residual Heat Removal System (PRHRS), and Containment Pressure and Radioactivity Suppression System (CPRSS) while i-SMR is equipped with the Passive Emergency Core Cooling System (PECCS), the Passive Containment Cooling System (PCCS), and the Passive Auxiliary Feedwater System (PAFS). These systems operate based on natural forces such as gravity and buoyancy, performing safety functions without external power or operator action. However, because their operation relies on relatively weak and time-varying driving forces, reliable modeling using system analysis codes is important. In particular, improper simulation of key thermal-hydraulic phenomena such as pressure drop, condensation, boiling, and natural circulation can lead to predictions that deviate considerably from actual performance. To address these concerns, this study reviews the modeling and simulation of PSIS, PAFS, and PCCS in Korean SMRs using various system analysis codes. Based on the authors' extensive experience, detailed modeling considerations are derived to improve the representation of key physical phenomena. Furthermore, the study discusses the importance of robustness evaluation under degraded conditions using the Best Estimate with Performance Issues (BEPI) framework. The insights provided herein are expected to support the credible and technically robust application of system analysis codes to the design and safety assessment of passive safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114747"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981786","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}

{"title":"Computational modeling of graphite degradation in molten salt reactors: Role of infiltration","authors":"Veerappan Prithivirajan ,&nbsp;Benjamin Spencer ,&nbsp;Joseph Bass ,&nbsp;Somayajulu L.N. Dhulipala ,&nbsp;Daniel Schwen ,&nbsp;Mustafa K. Jaradat","doi":"10.1016/j.nucengdes.2026.114796","DOIUrl":"10.1016/j.nucengdes.2026.114796","url":null,"abstract":"<div><div>Molten salt reactors (MSRs) often employ graphite as a moderator and reflector. An important challenge for deploying graphite in these reactors is that, due to limited experimental data, our understanding of graphite’s structural integrity in molten salt environments remains incomplete. This study addresses heat generation from fuel-bearing salt that has infiltrated open pores in the graphite, driven primarily by pressure differentials. This is one of multiple identified physical and chemical mechanisms through which molten salt could potentially degrade graphite. Thermally driven stresses are quantified using the Molten-Salt Reactor Experiment (MSRE) graphite moderator elements as a case study. Finite element simulations predict stress distributions at varying infiltration levels, indicating that thermal stresses increase with higher infiltration. Rare-event simulations using the parallel subset simulation framework identify the combinations and corresponding ranges of input parameters that lead to stresses above a specified threshold. In particular, combinations involving high infiltration amounts, high power density, and low thermal conductivity tend to induce the highest stresses. Under the inputs and assumptions considered in this work, the magnitudes of the thermally driven stresses are quite low, with a very low likelihood of causing failure due to exceeding the graphite’s tensile strength. Additionally, rare-event simulations were performed for two more scenarios: a scaled-up moderator geometry and a localized hotspot in the original geometry. Both cases resulted in increased susceptibility to failure, though not to a detrimental extent. Furthermore, the combined effects of irradiation and infiltration-induced thermal stresses were evaluated. The results showed that thermal stresses from infiltration were negligible compared to those caused by irradiation. The findings of such a study are inherently component-specific, but the methodology presented here could be used for similar assessments of salt-infiltration effects in other graphite components.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114796"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146057606","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}

{"title":"Design of nuclear fuel loading patterns for a PWR with Wasserstein generative adversarial networks","authors":"Anderson Alvarenga de Moura Meneses ,&nbsp;Lenilson Moreira Araujo","doi":"10.1016/j.nucengdes.2026.114765","DOIUrl":"10.1016/j.nucengdes.2026.114765","url":null,"abstract":"<div><div>The Loading Pattern (LP) design is part of the nuclear fuel management of a Nuclear Power Plant (NPP). The design of an LP includes the permutation of fuel assemblies, as well as calculations performed with reactor physics codes, aiming to producing energy with the satisfaction of constraints such as those related to safety. From a computational perspective, it is an NP-hard combinatorial problem solved with success by optimization metaheuristics. With the breakthrough of generative Artificial Intelligence (AI), the immediate question is whether LPs can be designed within such paradigm. In the present article, a methodology is proposed for training and applying Wasserstein Generative Adversarial Networks (WGANs) for automatic generation of LPs of a Pressurized Water Reactor. With the application of the methodology to the benchmark IAEA-2D, WGANs generated LPs satisfying the safety constraints and objectives proposed. Thus, WGANs can learn implicit probability distributions of nucler fuel and automatically design high-quality LPs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114765"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036031","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}

{"title":"Optimizing iPWR SMR core design: a power peaking factor analysis of annular fuel rods using MCNP5","authors":"Fatima Ghandour ,&nbsp;Salah Hamieh ,&nbsp;Ziad Francis","doi":"10.1016/j.nucengdes.2026.114768","DOIUrl":"10.1016/j.nucengdes.2026.114768","url":null,"abstract":"<div><div>This study investigates the neutronic performance of dual cooled annular fuel rods in the CAREM 25 integral Pressurized Water Reactor (iPWR), a small modular reactor (SMR), using MCNP5 Monte Carlo simulations. The motivation is to reduce power peaking factors (PPFs) and enhance thermal-hydraulic safety margins by adopting an annular fuel geometry with internal and external cooling. Three annular fuel configurations with 100%, 95%, and 93% fuel loading were analyzed and compared to the conventional solid fuel design. Geometric transformations were performed analytically—introducing, for the first time, closed-form equations for the inner and outer radii of annular fuel rods—to maintain the fuel-to-coolant volume ratio while limiting fuel mass reduction to ≤10%. The results show that the total PPF decreased by up to 27.45% in the 95% fuel loading case, dropping from 2.404 (solid design) to 1.744. Additionally, the effective multiplication factor (Keff) was reduced from 1.12445 to 1.09512, enhancing reactor controllability. The 95% loading configuration emerged as the optimal design, balancing neutronic performance and safety. These findings demonstrate that annular fuel can significantly flatten the power distribution and improve the safety profile of iPWR SMRs without compromising core performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114768"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036073","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}

{"title":"The impact of Rudi J. J. Stamm'ler on the development of the nuclear industry in Argentina","authors":"Eduardo Villarino ,&nbsp;Aldo Ferri","doi":"10.1016/j.nucengdes.2026.114817","DOIUrl":"10.1016/j.nucengdes.2026.114817","url":null,"abstract":"<div><div>This paper examines the technical and methodological contributions of Rudi J. J. Stamm'ler that shaped the historical development of nuclear technology and reactor physics capabilities in Argentina. Through successive missions supported by the <span>International Atomic Energy Agency</span> (IAEA), Rudi Stamm'ler played a decisive role in the training of highly qualified human resources at the <span>Balseiro Institute</span> and in the establishment of national computational capabilities in reactor physics and design. His influence extended to the development of essential calculation tools, the consolidation of independent nuclear design methodologies in Argentina, and the professional growth of engineers who later contributed to major national and international nuclear initiatives. The long-term impact of this work is reflected in the continued use of the computational methods he promoted and in an enduring technical and academic legacy within Argentina's nuclear engineering community.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114817"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190254","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}

{"title":"Experimental investigation of gas-phase migration behaviors in a 1 × 2 rod bundle under two-phase equilibrium and non-equilibrium flow","authors":"Hao Xie,&nbsp;Wenhai Qu,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2026.114806","DOIUrl":"10.1016/j.nucengdes.2026.114806","url":null,"abstract":"<div><div>Void fraction gradient was assumed as the mechanism of void drift between subchannels. However, classical void drift models based on this assumption predict poor results under bubbly flow and cap-bubbly flow against experimental data. In this work, void fraction and bubble diameter in an enlarged 1 × 2 rod bundle was experimentally investigated by wire mesh sensors (WMS) system under equilibrium and non-equilibrium gas flow rates. The total 168 cases cover bubbly flow, cap-bubbly flow and churn flow defined with bubble shape and volume-base probability density function (VPDF). In general, small bubbles gather near channel walls, while large bubbles concentrate in sub-channel centers. For bubbly flow and cap-bubbly flow, it is difficult for bubbles passing through gap between subchannels. Thus, classical void drift models deviate from experimental results in bubbly flow and cap-bubbly flow. However, when bubble diameter is larger than 0.75 time of pitch of rod bundle, void drift improves obviously because bubbles are large enough to disturb the corresponding subchannel. Under bubbly flow and cap-bubbly flow, large bubbles coalesce in each subchannel. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. For churn flow, strong void fraction migration between sub-channels is mainly caused by large bubbles with diameter larger than pitch of 32 mm. Thus, equilibrium distribution of void fraction can be fully developed within distance of 1830 mm under equilibrium and non-equilibrium inlet conditions. This is the reason that classical void drift models predict well against experimental data in churn flow. Large bubbles break down in churn flow. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. This work helps to under mechanism of gas migration between sub-channels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114806"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190735","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}