Nuclear Engineering and Design最新文献

{"title":"Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT","authors":"Jacob A. Hirschhorn,&nbsp;Mustafa K. Jaradat,&nbsp;Ryan T. Sweet,&nbsp;Nicolas E. Woolstenhulme,&nbsp;Paul A. Demkowicz,&nbsp;David A. Reger,&nbsp;Paolo Balestra,&nbsp;Gerhard Strydom","doi":"10.1016/j.nucengdes.2026.114773","DOIUrl":"10.1016/j.nucengdes.2026.114773","url":null,"abstract":"<div><div>The United States nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles’ silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114773"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036058","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":"Side-by-side high-temperature accident performance of ATF and conventional claddings in the CODEX-ATF experiment","authors":"Nóra Vér ,&nbsp;Róbert Farkas ,&nbsp;Berta Bürger ,&nbsp;Anna Pintér-Csordás ,&nbsp;Tamás Novotny ,&nbsp;Erzsébet Perez-Feró ,&nbsp;Péter Szabó ,&nbsp;Levente Illés ,&nbsp;Zoltán Kovács ,&nbsp;Dávid Cinger ,&nbsp;Martin Ševeček ,&nbsp;Zoltán Hózer","doi":"10.1016/j.nucengdes.2026.114770","DOIUrl":"10.1016/j.nucengdes.2026.114770","url":null,"abstract":"<div><div>The CODEX-ATF integral bundle test was conducted within the framework of the IAEA Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS) project at the CODEX (COre Degradation Experiment) facility in Hungary. The electrically heated seven-rod bundle consisted of four Cr-coated and three uncoated Zr alloy cladding tubes, enabling a direct comparison of their behavior under high-temperature accident conditions. The experiment primarily aimed to investigate fuel failure and degradation mechanisms.</div><div>During the test, several rods exhibited ballooning and burst phenomena. The maximum temperature exceeded 1600 °C. The transient was terminated by a bottom-up water quench. The total hydrogen generation was approximately 3 g, indicating substantial oxidation of the zirconium-based components. Intensive Zr-Cr eutectic interaction was observed in the hottest region of the bundle on the Cr-coated claddings. Post-test examinations revealed pronounced deformation and failure in both coated and uncoated claddings.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114770"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081254","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":"Multi-Physics Benchmark for a Thermal Molten Salt Reactor","authors":"P. Pfahl ,&nbsp;B. Kędzierska ,&nbsp;I. Kim ,&nbsp;A. Chambon ,&nbsp;L. Fischer ,&nbsp;L. Bureš ,&nbsp;J. Groth-Jensen ,&nbsp;Y. Kim ,&nbsp;A. Rineiski ,&nbsp;B. Lauritzen","doi":"10.1016/j.nucengdes.2026.114790","DOIUrl":"10.1016/j.nucengdes.2026.114790","url":null,"abstract":"<div><div>Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114790"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078926","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":"A state-of-the-art review of R&D for the supercritical water-cooled reactor technology. Part II materials & chemistry","authors":"K. Khumsa-Ang ,&nbsp;M. Fulger ,&nbsp;A. Sáez Maderuelo ,&nbsp;A. Toivonen ,&nbsp;M. Sipova ,&nbsp;D. Marusakova ,&nbsp;L. Zhang ,&nbsp;J. Macák ,&nbsp;R. Novotny","doi":"10.1016/j.nucengdes.2026.114789","DOIUrl":"10.1016/j.nucengdes.2026.114789","url":null,"abstract":"<div><div>This document presents a summary of the most relevant research and development (R&amp;D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&amp;D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).</div><div>One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114789"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078928","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":"Prediction of leakage rates from containment buildings under ultimate internal pressure","authors":"Jin-Young Park,&nbsp;Tae-Hyun Kwon","doi":"10.1016/j.nucengdes.2026.114792","DOIUrl":"10.1016/j.nucengdes.2026.114792","url":null,"abstract":"<div><div>Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114792"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078930","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":"Integrated multiphysics assessment of grid-to-rod fretting mitigation: enhancing nuclear safety through improved fuel integrity","authors":"Quazi Md. Zobaer Shah ,&nbsp;Md Mahabub Hasan Mousum ,&nbsp;Debashis Datta ,&nbsp;Md. Arefin Kowser ,&nbsp;Mohammad Asaduzzaman Chowdhury ,&nbsp;Md. Shojib Mia ,&nbsp;Shahpara Sheikh Dola","doi":"10.1016/j.nucengdes.2025.114696","DOIUrl":"10.1016/j.nucengdes.2025.114696","url":null,"abstract":"<div><div>Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114696"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950226","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":"PIV study on subchannel cross-flow characteristics in 19-pin wire-wrapped bundle channels","authors":"Rongjie Li ,&nbsp;Wukun Zhu ,&nbsp;Chengwen Qiang ,&nbsp;Minghan He ,&nbsp;Han Wang ,&nbsp;Lu Liu ,&nbsp;Sicheng Wang ,&nbsp;Dajun Fan","doi":"10.1016/j.nucengdes.2026.114810","DOIUrl":"10.1016/j.nucengdes.2026.114810","url":null,"abstract":"<div><div>The China Initiative Accelerator Driven System (CiADS) holds significant promise in the transmutation of long-lived nuclear waste and the efficient utilization of nuclear fuel. Understanding the coolant flow characteristics within reactor fuel assemblies is crucial for ensuring the safe and economical operation of nuclear facilities. In this study, a visual hydraulic experimental platform was constructed to investigate the flow velocity distribution within the fuel assembly of CiADS. Utilizing particle image velocimetry technology, the transverse velocity distribution at subchannel interfaces within a 19-pin wire-wrapped fuel assembly was systematically investigated. The results reveal a distinct cosine-type distribution of the normalized transverse velocity at the interfaces between internal subchannels, with key flow features demonstrating robust Reynolds-number independence. Notably, significant wall-induced asymmetries alter this distribution at the interface between the internal subchannel and the side-wall subchannel. A novel transverse mixing model is developed, which shows close agreement with experimental data, confirming its predictive capability for inter-subchannel mixing in wire-wrapped bundles. This research further confirms the periodic disturbance characteristics induced by wire-wraps on transverse flow and provides a theoretical basis for predicting normalized transverse velocity in wire-wrapped bundles of different sizes, thereby deepening the understanding of the transverse mixing mechanism within rod bundles.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114810"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190730","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":"CFD simulation methodology for large-scale heavy liquid metal facility relevant for MYRRHA: Application to E-SCAPE facility","authors":"S. Lopes ,&nbsp;S. Keijers ,&nbsp;L. Koloszar ,&nbsp;J. Pacio ,&nbsp;P. Planquart ,&nbsp;K. Van Tichelen","doi":"10.1016/j.nucengdes.2026.114798","DOIUrl":"10.1016/j.nucengdes.2026.114798","url":null,"abstract":"<div><div>This article is part of a series devoted to the analysis of pool thermal-hydraulic phenomena in heavy-liquid metal systems, combining experimental investigations with numerical simulations. The reference system is the Accelerator Driven System MYRRHA, currently under construction at SCK CEN (Belgium), which couples a particle accelerator with a pool-type research reactor cooled by lead-bismuth eutectic (LBE). An extensive experimental campaign has been carried out in the E-SCAPE facility, a 1/6-scaled LBE mockup of the primary vessel of MYRRHA, equipped with additional auxiliary circuits. In parallel, numerical studies based on computational fluid dynamics (CFD) are performed with the main objective of developing and validating robust CFD methodology capable of accurately simulating large-scale heavy-liquid metal experimental facilities. These studies allow the investigation of thermal-hydraulic phenomena and provide detailed insight into thermal and momentum fields. Model validation is achieved through systematic comparison between simulations and measurements obtained in E-SCAPE for a range of operating scenarios, including forced and natural circulation, symmetric and asymmetric loss-of-flow events, and asymmetric decay heat removal conditions.</div><div>In this paper, the numerical methodology to simulate the E-SCAPE facility under both steady-state and transient operating conditions is established. Since the natural circulation condition and the transition from forced to natural circulation are characterized by a strong coupling between thermal and momentum fields, detailed numerical description of the E-SCAPE core simulator, of the upper plena regions and of the external circuits are important. The main reason is that the flow rate in the core region is driven by the balance between the pressure drop and the thermal buoyancy effect and the flow rates in the external circuits are based on pressure losses through the valves, filters, and pumps. The impact of geometric simplifications of core plates and heaters, relative to the real configuration, is evaluated on the flow and on the temperature fields.</div><div>Furthermore, the effects of numerical mesh refinement and thermal boundary conditions are investigated under both iso-thermal and thermal forced circulation operating conditions. Simulated temperature distributions and pressure losses in the pool are compared against local experimental measurements. Finally, practical guidelines are proposed for CFD simulation of forced circulation, natural circulation, and transient conditions, addressing geometry simplifications, definition of fluid and solid domains, characterization of external circuits, and numerical strategies for transient initiation. The initial conditions for transient simulations are also computed and successfully compared with experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114798"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190733","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":"Erratum to “An analysis of the brittleness indices of SiC layer in the TRISO fuel,” Nucl. Eng. Design 446 (2026) 114614.","authors":"Makuteswara Srinivasan","doi":"10.1016/j.nucengdes.2026.114793","DOIUrl":"10.1016/j.nucengdes.2026.114793","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114793"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189363","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":"Thermo-mechanical coupled analysis and probabilistic safety evaluation of the prestressed concrete containment vessel under accident pressure","authors":"Hua Rong ,&nbsp;Xuan Zhang ,&nbsp;Yajing Shen ,&nbsp;Shuai Tan ,&nbsp;Xinglang Fan ,&nbsp;Yang Du ,&nbsp;Yifeng Feng","doi":"10.1016/j.nucengdes.2026.114760","DOIUrl":"10.1016/j.nucengdes.2026.114760","url":null,"abstract":"<div><div>This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114760"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078929","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}