Nuclear Engineering and Design最新文献

{"title":"Subchannel thermal–hydraulic analysis of (UN + U3Si2)-SiC system considering Chalk River Unidentified deposits layer","authors":"Mingdong Kai ,&nbsp;Jiejin Cai","doi":"10.1016/j.nucengdes.2025.114126","DOIUrl":"10.1016/j.nucengdes.2025.114126","url":null,"abstract":"<div><div>The (UN + U₃Si₂)-SiC fuel-cladding combination is one of the most potential accident-tolerant fuel-cladding combinations. In this paper, the influence of surface fouling on core heat transfer is considered, and the thermal–hydraulic characteristics of the accident tolerant fuel (UN + U₃Si₂)-SiC fuel-cladding combination under steady-state conditions and typical LOFA accident conditions are analyzed. Firstly, the model constructed in this paper is systematically validated through PSBT(PWR Subchannel and Bundle Test) benchmark experiments. Then the variation of thermal parameters under steady-state and transient operating conditions when CRUD (Chalk River Unidentified Deposits) is present and at different thicknesses is further explored by analyzing the thermophysical properties of the (UN + U₃Si₂)-SiC fuel-cladding combination. Finally, the safety criterion parameters considering cladding corrosion during LOFA accident are analyzed, and the core safety is judged. The results show that the presence of CRUD will increase the MFCT and MCT of the reactor core and deteriorate the heat transfer. And the thicker the CRUD, the higher the degree of heat transfer deterioration. The existence of CRUD will also reduce the core safe operation parameter MDNBR (Minimal departure from nuclear boiling ratio), but due to the superior performance of the accident tolerant fuel (UN + U₃Si₂)-SiC fuel-cladding combination, the impact of CRUD on the core safe operation is still within the controllable range.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114126"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916425","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":"Chemical interactions between La(iii) and O2− in molten fluoride salts and La(iii)/U(iv) separation by oxide precipitation method","authors":"Hao Peng ,&nbsp;Nan Ji ,&nbsp;Bo Zhou ,&nbsp;Yunlong Wei ,&nbsp;Wei Huang ,&nbsp;Yu Gong","doi":"10.1016/j.nucengdes.2025.114127","DOIUrl":"10.1016/j.nucengdes.2025.114127","url":null,"abstract":"<div><div>The chemical interactions between La(<span>iii</span>) and O²⁻ in 66.7LiF-33.3BeF₂ (FLiBe) and 46.5LiF-11.5NaF-42KF (FLiNaK) molten salt systems at 873 K were studied by dissolution and oxide titration methods. In the FLiBe, the precipitation-dissolution behavior of La₂O₃ is a simple equilibrium mechanism between La(<span>iii</span>) and O²⁻ ions. The solubility of La₂O₃ in FLiBe melt was 0.078 mol/kg with the dissolution equilibrium time of 5 h, and the corresponding apparent solubility product (<span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>) of La₂O₃ was (3.43 ± 0.75) × 10⁻⁴ mol⁵/kg⁵. The oxide titration experiment showed that the product of the interaction between La(<span>iii</span>) and O²⁻ in FLiBe is La₂O₃ precipitate, and the <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span> was (3.45 ± 0.37) × 10⁻⁴ mol⁵/kg⁵, which was highly consistent with that obtained by the dissolution method. Based on the <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span> value, the oxide tolerance for La₂O₃ precipitation was then evaluated. However, the chemical reaction between La(<span>iii</span>) and O²⁻ in FLiNaK was more complicated. The dissolution of La₂O₃ would produce oxyfluoride LaOF, and addition of Li₂O into the FLiNaK-La(<span>iii</span>) molten salt could cause precipitation of equimolar solid compounds La₂O₃ and LaOF. The oxyfluoride species LaOF was correlated with a high content of free fluoride ions (F⁻) in FLiNaK. At last, an oxide precipitation method was proposed for La(<span>iii</span>)/U(<span>iv</span>) separation based on the analysis of <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>(La₂O₃) and <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>(UO₂), and this method achieved a good La(<span>iii</span>)/U(<span>iv</span>) separation efficiency in the FLiBe-LaF₃-UF₄ melt.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114127"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913236","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":"Material properties and models for the assessment of pressure boundary failure in high-pressure core melt accident scenarios","authors":"Christoph Bläsius ,&nbsp;Jürgen Sievers ,&nbsp;Annett Udoh ,&nbsp;Stefan Weihe","doi":"10.1016/j.nucengdes.2025.114109","DOIUrl":"10.1016/j.nucengdes.2025.114109","url":null,"abstract":"<div><div>The simulation of pressure boundary failure in high-pressure core melt accident scenarios requires specific material data, which are often rare. This work focuses on the behavior of German steel grades and alloys used for KWU-built reactors in this extreme load range. In a first step, relevant materials are described. Previous work is presented in brief. A series of experimental studies addresses open questions regarding the material behavior in specific situations and for specific components, such as short-term creep of Alloy 800 (mod.), tearing behavior of 20 MnMoNi 5 5, relaxation behavior of bolt steels, influence of high-temperature oxidation on fracture, and behavior of contact surfaces in safety and relief valves. Finally, material models for the major steel grades are presented, verified, validated and their accuracy and associated uncertainties are discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114109"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913239","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 the spectral element CFD solver on Sunway TaihuLight for nuclear reactor simulation","authors":"Lingyu Dong,&nbsp;Zhifeng Zhou,&nbsp;Genshen Chu,&nbsp;Dandan Chen,&nbsp;Hongzhen Zhang,&nbsp;Yang Li","doi":"10.1016/j.nucengdes.2025.114071","DOIUrl":"10.1016/j.nucengdes.2025.114071","url":null,"abstract":"<div><div>High-fidelity computational fluid dynamics (CFD) plays a crucial role in analyzing thermal–hydraulic phenomena in advanced nuclear reactors. This study presents an optimization of the spectral element method (SEM)-based CFD solver Phiflow-Solver on Sunway TaihuLight supercomputer to accelerate nuclear reactor simulations. The SEM solver relies on small, dense matrix multiplications and the Poisson operator, which are computationally challenging on heterogeneous architectures. To address these challenges, we propose two optimization strategies: (1) Porting matrix operations to the SW26010 processor’s Computing Processing Elements (CPEs) using DMA-enhanced data transfer and SIMD vectorization, achieving a 51.9% performance improvement at 64 CGs for a polynomial order of 24; (2) Enabling collaborative Management Processing Element (MPE)-CPE parallelism to compute multiple spectral elements simultaneously, achieving a 65.5% performance gain under identical conditions. By integrating these strategies, we achieve an overall 70.6% performance enhancement. Validation with a 7-pin wire-wrapped fuel assembly confirms that the heterogeneous optimizations maintain the solver’s accuracy. Furthermore, as the mesh size scales from 42 million to 1.3 billion grid points, the weak scalability remains above 90%, demonstrating the solver’s improved capability for high-resolution nuclear fuel assembly simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114071"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916424","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":"Development of an evaluation method for debris bed formation behavior focusing on the agglomeration mechanism observed in the DEFOR-A test using THERMOS/JBREAK–DPCOOL–MSPREAD","authors":"Wataru Kikuchi ,&nbsp;Akitoshi Hotta ,&nbsp;Koetsu Ito ,&nbsp;Mamoru Shimizu","doi":"10.1016/j.nucengdes.2025.114041","DOIUrl":"10.1016/j.nucengdes.2025.114041","url":null,"abstract":"<div><div>When the lower head of a reactor pressure vessel (RPV) is damaged during a severe accident in light water reactors (LWRs), after the jet breakup occurs in the water, the entrained pieces of molten debris (hereafter called droplets) is likely to form nonuniform, poorly coolable agglomerations on the floor. These debris agglomerates can impact the debris bed coolability. The authors are developing THERMOS, an analysis code composed of modules such as JBREAK, DPCOOL, and MSPREAD, to evaluate these behaviors. In the investigation of the new DEFOR-A test series conducted in collaboration with Kungliga Tekniska Högskolan (KTH), it has been identified that the formation of agglomerated debris is influenced not only by the solidification fraction of the droplets but also by crust cracking and melt spreading. To evaluate the formation of agglomerated debris at a wide range of superheat, the authors have developed the special model in JBREAK, one of the THERMOS modules, based on mechanism estimated from the investigation of DEFOR-A test series (A23-27). Additionally, the agglomeration process is affected by several complex phenomena, such as jet breakup, droplet sedimentation, deposition, and melt spreading behavior, so the authors developed an evaluation method that sequentially evaluates these behaviors using the THERMOS/JBREAK–DPCOOL–MSPREAD coupling. This evaluation method successfully simulated jet breakup, agglomeration, and debris bed formation observed in the DEFOR-A tests. The evaluation method has accurately explained the agglomerated debris mass fraction over a wide range of melt superheat levels by modeling droplet crust cracking, melt spreading, and agglomeration resulting from droplet–debris interactions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114041"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913237","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 progress in high-temperature thermo-mechanical behaviors for modelling Cr-coated cladding under loss-of-coolant accident condition","authors":"Chunyu Yin ,&nbsp;Guanghui Su ,&nbsp;Libo Qian ,&nbsp;Qingwen Xiong ,&nbsp;Yu Liu ,&nbsp;Yingwei Wu ,&nbsp;Sijia Du ,&nbsp;Jing Zhang ,&nbsp;Zhong Xiao","doi":"10.1016/j.nucengdes.2025.114125","DOIUrl":"10.1016/j.nucengdes.2025.114125","url":null,"abstract":"<div><div>Chromium (Cr)-coated zirconium cladding has emerged as a leading candidate for accident tolerant fuel (ATF) cladding in near-term engineering applications. This cladding demonstrates enhanced resistance to high-temperature oxidation, superior mechanical properties at elevated temperatures, and a relatively high level of technological maturity. Its performance under loss-of-coolant accident (LOCA) conditions is critical to reactor safety, making it a key focus of the present study. The present work introduces an overview of research progress on high temperature thermo-mechanical behaviors for Cr-coated cladding and provides a set of fundamental safety analysis models tailored for LOCA scenarios. First, essential models for LOCA safety analysis of Cr-coated cladding are identified, including a high-temperature oxidation model (along with a Cr coating consumption model), a high-temperature creep model, a high-temperature burst model, and an embrittlement criterion. Second, based on the evaluation of experimental data from high-temperature oxidation studies, models for the growth of Cr₂O₃ layer and oxygen absorption are recommended to estimate the oxidation rate of Cr-coated cladding. Additionally, a model for Cr coating consumption is proposed. Subsequently, through a comprehensive review and reevaluation of high-temperature creep and burst data, corresponding models for Cr-coated cladding are developed respectively. Finally, embrittlement data for Cr-coated cladding are analyzed, and embrittlement criteria for both one-sided oxidation and two-sided oxidation conditions are proposed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114125"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913238","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 of interassembly bypass flow in Sodium Fast Reactors","authors":"A.J. Novak ,&nbsp;C. Bourdot Dutra ,&nbsp;D. Shaver ,&nbsp;E. Merzari","doi":"10.1016/j.nucengdes.2025.114044","DOIUrl":"10.1016/j.nucengdes.2025.114044","url":null,"abstract":"<div><div>Interassembly flow in Sodium Fast Reactors (SFRs) represents a bypass flow path exterior to the fuel assembly ducts. Heat transferred across this thin gap is an important component of core radial expansion, where the coupling between thermal-fluids, neutronics, and solid mechanics results in time-dependent duct bowing. These geometry changes can constitute a significant portion of the total reactivity response in transients, but are difficult to model in high-fidelity. Interassembly flow is also an important heat transfer mode during natural convection cooling. To improve our understanding of interassembly flow, this paper provides NekRS Reynolds Averaged Navier–Stokes (RANS) and Large Eddy Simulations (LES) of the interassembly flow in a 19-bundle fast reactor core. Time-averaged LES compares reasonably well with a <span><math><mi>k</mi></math></span>-<span><math><mi>τ</mi></math></span> RANS model, though RANS is not able to capture a crossflow which occurs at a large change in flow area between the duct–duct gaps and the open peripheral region. We predict velocity distributions and illustrate a multiscale postprocessing system that can be used to generate coarse-mesh closures for subchannel and porous media tools, and provide a dataset with average velocity for comparison with coarse-mesh tools.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114044"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906224","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 study on post-buckling behavior of fast reactor vessel under excessive earthquakes","authors":"Yiji Ye,&nbsp;Masakazu Ichimiya,&nbsp;Naoto Kasahara","doi":"10.1016/j.nucengdes.2025.114130","DOIUrl":"10.1016/j.nucengdes.2025.114130","url":null,"abstract":"<div><div>The Fukushima Daiichi nuclear accident has raised the nuclear industry’s interest in the countermeasures for Beyond Design Basis Events (BDBEs) such as excessive earthquake. Sable failure modes are acceptable in BDBEs with the safety goal being to prevent unstable failure modes. The Fast Reactor Vessel (FRV) is vulnerable to seismic buckling due to thin-walled structure. Under excessive earthquakes, the safety goal of FRV is to achieve a stable post-buckling state. This paper presents an experimental study on post-buckling behavior of short and medium cylinders under horizontal vibration, simulating the phenomena in pool and loop type FRV. Independent of buckling configuration, a global response stability is confirmed after buckling. This stability is achieved by the phase-shift phenomenon, where buckling initiation increases the frequency ratio and enables the displacement-controlled characteristic of the dynamic load. Such phenomenon is independent of input conditions. In addition, longer cylinders show a higher post-buckling frequency ratio with significant response reduction compared to short cylinders. Next, the post-buckling failure development process is investigated and can be summarized into three stages. The ultimate rupture boundary can be measured by the critical cumulative input energy, which shows clear dependency on the buckling configuration. A preliminary criterion against ultimate rupture and the energy-based failure mode map are proposed to assess the safety margin of FRV. It demonstrates a considerable margin from buckling initiation to ultimate rupture during an excessive earthquake. This paper largely extends the database and contributes to a more comprehensive understanding in the post-buckling domain of FRV under BDBEs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114130"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906316","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":"Optimized ensemble of neural networks for the prediction of critical heat flux","authors":"Ibrahim Ahmed ,&nbsp;Irene Gatti ,&nbsp;Enrico Zio","doi":"10.1016/j.nucengdes.2025.114111","DOIUrl":"10.1016/j.nucengdes.2025.114111","url":null,"abstract":"<div><div>Critical Heat Flux (CHF) is a thermal limit in boiling heat transfer, beyond which there is a substantial reduction in heat transfer efficiency. This phenomenon plays a vital role in the thermal engineering design of systems involving two-phase flow. As a result, an accurate CHF prediction is essential for both safety and performance, particularly in water-cooled nuclear reactors where thermohydraulic margins are critical. In this paper, a novel optimized ensemble of neural networks (NNs) for CHF prediction is proposed to enhance the accuracy of individual models trained separately with distinct architectures and hyperparameters settings. Two systematic procedures are presented to identify potentially optimal NN models and aggregate them into an optimal ensemble model. The proposed method is validated using experimental CHF data made available by the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) Expert Group on Reactor Systems Multi-Physics (EGMUP) task force on AI and ML for Scientific Computing in Nuclear Engineering projects, promoted by the OECD/NEA. The results obtained show that the ensemble model outperforms standalone models and other state-of-the-art modelling approaches. Parametric and sensitivity analyses across various input parameters confirm the robustness of the ensemble model and its consistency with expected physical behaviors, further underlying its potential for improving CHF prediction in nuclear reactor applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114111"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906225","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":"Radiation safety and fuel performance of pebble-bed modular high-temperature gas-cooled reactors","authors":"Yu Wang ,&nbsp;Jianzhu Cao ,&nbsp;Bing Xia ,&nbsp;Feng Xie ,&nbsp;Fu Li ,&nbsp;Haitao Wang ,&nbsp;Jiejuan Tong ,&nbsp;Yujie Dong ,&nbsp;Zuoyi Zhang ,&nbsp;Karl Verfondern ,&nbsp;Sudarshan K. Loyalka","doi":"10.1016/j.nucengdes.2025.114116","DOIUrl":"10.1016/j.nucengdes.2025.114116","url":null,"abstract":"<div><div>Safety of nuclear reactors is of wide concern in the world, and inherent safety of the reactors is the goal that the nuclear energy field has been pursuing over the last several decades. How to quantitatively evaluate the inherent safety, as well as the reactor radiation safety, is a long-term important scientific and technical issue. This study focused on the 10 MW high-temperature gas-cooled experimental reactor (HTR-10), the only operational pebble-bed modular HTGR for testing that can operate at full power currently, measured its primary coolant activity, a key indicator of reactor radiation safety, and assessed its fuel element performance, directly affecting its inherent safety feature. A method for evaluating tri-structural isotropic coated fuel particle (TRISO CFP) failure fraction and uranium contamination share was established. The release-to-birth (R/B) ratio for fission gas nuclides was &lt; 1 × 10⁻⁶, with a TRISO CFP failure fraction of 7.23 × 10⁻⁵ and uranium contamination share of 9.20 × 10⁻⁶. The TRISO CFP performance of HTR-10 surpassed that of previous HTGRs and irradiation tests conducted in the world, highlighting its excellent radiation safety and potential for large-scale commercial application of HTR-PM (that are related/based on HTR-10).</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114116"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906317","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}