Annals of Nuclear Energy最新文献

{"title":"Energy-dependent neutron removal cross-section","authors":"A. Soliman","doi":"10.1016/j.anucene.2024.111171","DOIUrl":"10.1016/j.anucene.2024.111171","url":null,"abstract":"<div><div>This study provides new, essential data and models for the monoenergetic neutron removal cross-section required for neutron shielding. A Monte Carlo model using CERN FLUKA code was developed to calculate the energy-dependent removal cross-section for elements commonly used in radiation shielding materials. Neutron energies ranging from 0.5 to 20 MeV including the fusion neutrons at 2.5 and 14.1 MeV were analyzed to generate the required data. Unlike previous approaches, which often rely on approximations or limited datasets, this study provides comprehensive and precise models that describe energy-dependent removal cross-sections as functions of atomic weight and total neutron cross-section. The new models were successfully applied to neutron energies of 2.5 MeV and above, with enhanced accuracy for fission neutrons. The calculated removal cross-section data were compared with prior measurements and calculations, demonstrating the model’s capability to accurately predict the energy-dependent removal cross-sections for shielding materials such as iron, graphite, aluminum, lead, barite concrete, borosilicate glass, and polyethylene.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111171"},"PeriodicalIF":1.9,"publicationDate":"2025-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163629","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 new and rigorous SPN theory – The concluding Part V: Completion of the numerical solution method for GSP3 equations","authors":"Yung-An Chao ,&nbsp;Ru-Ying Tang ,&nbsp;Ziyong Li ,&nbsp;Lianghui Peng","doi":"10.1016/j.anucene.2024.111169","DOIUrl":"10.1016/j.anucene.2024.111169","url":null,"abstract":"<div><div>We published a series of four papers on the development of the Generalized SP_N (GSP_N) theory. The theory contains only diffusion equations and composes of different levels of approximation, <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mi>N</mi></mrow><mrow><mo>(</mo><mi>K</mi><mo>)</mo></mrow></msubsup></mrow></math></span>. Its highest level <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mi>N</mi></mrow><mrow><mo>(</mo><mi>N</mi><mo>)</mo></mrow></msubsup></mrow></math></span> is equivalent to P_N, while its lowest level <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mi>N</mi></mrow><mrow><mo>(</mo><mn>0</mn><mo>)</mo></mrow></msubsup></mrow></math></span> has the same differential equations as the traditional SP_N but with different interface and boundary conditions. These conditions contain additional tangential derivatives on the surfaces, which poses a serious challenge to traditional methods of numerical calculation. In Part IV of the series, we introduced the generalized transverse integration nodal (GTIN) method to solve the <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mn>3</mn></mrow><mrow><mo>(</mo><mn>0</mn><mo>)</mo></mrow></msubsup></mrow></math></span> equations. Despite the very encouraging numerical results, there was the problem that the base functions could satisfy only part, not all, of the interface and boundary conditions. In this concluding part of the series, we resolve this last issue to complete the GTIN method for numerically solving the <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mn>3</mn></mrow><mrow><mo>(</mo><mn>0</mn><mo>)</mo></mrow></msubsup></mrow></math></span> equations. Higher order base functions are derived that can provide additional parabolically weighted surface moments in order to meet all the required interface and boundary conditions. However, we discovered the fundamental problem that the presence of second order tangential derivatives in the interface/boundary condition means the condition itself being a differential equation with two degrees of freedom such that the solution is not unique but infinitely many. This issue is successfully resolved so that we can calculate the physical solution very easily despite all the possible mathematical solutions. Numerical results confirm the theoretical analysis and expectation, and demonstrate the better accuracy of <span><math><mrow><mi>G</mi><mi>S</mi><msubsup><mi>P</mi><mrow><mn>3</mn></mrow><mrow><mo>(</mo><mn>0</mn><mo>)</mo></mrow></msubsup></mrow></math></span> versus SP₃. The method is potentially applicable to the generic geometry of arbitrary polygon. An Errata is included for correction to typos and minor errors in previous parts of this serial.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111169"},"PeriodicalIF":1.9,"publicationDate":"2025-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163583","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 a two-phase flow solver with drift-flux model based on OpenFOAM: Validation against single/two-phase and boiling flow","authors":"Wenqiang Wu ,&nbsp;Tao Huang ,&nbsp;Peng Du ,&nbsp;Dalin Zhang ,&nbsp;Lei Zhou ,&nbsp;Bo Wang ,&nbsp;Jian Deng ,&nbsp;Zhifang Qiu ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu ,&nbsp;Guanghui Su","doi":"10.1016/j.anucene.2024.111179","DOIUrl":"10.1016/j.anucene.2024.111179","url":null,"abstract":"<div><div>Based on the open-source finite volume method platform OpenFOAM, this paper implements a two-phase flow solver for drift partial non-equilibrium models with realistic closure relations. The solver inherits the numerical method from the OpenFOAM framework, adopts the large-time step transient PIMPLE algorithm, and follows its discrete and matrix-solving criteria, focusing on developing a two-phase flow model and constitutive models. The constitutive model is utilized as a library by the drift flux solver, covering the pre-CHF flow patterns in vertical channels. The drift velocity model is based on Ishii and his collaborators’ work. The remaining models, such as wall friction, wall heat transfer, interfacial heat and mass transfer, are taken from TRACE Theory Manual. To fulfill the condition of conjugate heat transfer between fluid and solid, a virtual solid heat transfer function is developed under the premise of a single-domain solver. Extensive and successful validation has been conducted involving single-phase, two-phase, and boiling flow heat transfer phenomena by comparing with experimental data or typical system codes, and the two-phase flow solver’s success has been demonstrated in several aspects. This will support future multi-scale and multi-physics coupling work within the OpenFOAM framework.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111179"},"PeriodicalIF":1.9,"publicationDate":"2025-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163627","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 in-vessel injection strategy for IVR improvement","authors":"Zheng Mingguang,&nbsp;Yan Jinquan,&nbsp;Cao Kemei,&nbsp;Lu Wei,&nbsp;Shi Guobao,&nbsp;Wang Jiayun,&nbsp;Zhang Kun","doi":"10.1016/j.anucene.2024.111155","DOIUrl":"10.1016/j.anucene.2024.111155","url":null,"abstract":"<div><div>In-vessel retention (IVR) of molten core debris is a significant severe accident management strategy to prevent vessel failure and subsequent debris relocation to the containment. High-risk ex-vessel phenomena such as steam explosion, core-concrete interaction (CCI) that may challenge the containment integrity are eliminated by successful IVR measures. In unmitigated severe accident, loss of cooling leads to core heat-up and molten debris relocates into the plenum of reactor vessel. High heat flux from molten pool to the vessel may melt through the wall, which depends on stratification of the debris and critical heat flux (CHF) of the ex − vessel surface. The debris configuration used to be postulated as a metal layer floating on the oxide layer. Nowadays thin metal layer over an oxide layer and bottom metal layer configuration is considered as the limiting and reasonable assumption. For high power nuclear power plant (NPP) with less metal in molten pool, if three-layer molten pool formed in the reactor vessel lower head, top thin metal layer may lead to focusing effect and challenge the integrity of vessel. In-vessel injection is recommended to mitigate focusing effect.CAP1400 is designed as passive GEN-III PWR and adopts IVR as main severe accident management strategy. For CAP1400 NPP IVR design, reactor cavity flooding system provides enough containment water level for natural circulation ex-vessel cooling, which is beneficial to water backflow from break. Emergency operating procedure (EOP) and severe accident management guideline (SAMG) also instruct plant staff recovering other engineering injection path to the reactor vessel manually. Therefore lower head debris decay heat removal is achieved by both in-vessel and ex-vessel cooling. Experiments show that debris top cooling heat transfer is enhanced compared to the one predicted by film boiling correlation. Heat transfer model was established to investigate the top cooling capability, including natural convection, radiation heat transfer, nucleate boiling and film boiling. Two classic three − layer molten pool scenarios were analyzed, according to the CAP1400 severe accident sequence and debris quality participated in interaction. Results show that some of the three − layer assumption cannot lead to vessel failure without in − vessel injection. For more limiting three − layer case, heat flux to the vessel wall decreases lower than CHF with conservative top cooling heat flux. Based on the test and deterministic assessment, there is reasonable assurance that the IVR strategy is successful by both in-vessel and ex-vessel cooling.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111155"},"PeriodicalIF":1.9,"publicationDate":"2024-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163626","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":"Numerical Investigation on molten Lead–Bismuth Eutectic solidification characteristics outside vertical tube bundles","authors":"Mengyu Zhu ,&nbsp;Zhen Wang ,&nbsp;Shichao Zhang ,&nbsp;Jiangtao Jia ,&nbsp;Shuyong Liu ,&nbsp;Taosheng Li","doi":"10.1016/j.anucene.2024.111172","DOIUrl":"10.1016/j.anucene.2024.111172","url":null,"abstract":"<div><div>The molten Lead–Bismuth Eutectic (LBE) solidification is a critical safety issue in the operation of lead-cooled fast reactors. Under overcooling conditions, the LBE located in the heat exchanger is prone to solidification. However, there are currently no studies on the LBE solidification outside the vertical tube bundles. In this paper, a three-dimensional CFD model of LBE solidification outside the vertical heat exchange tube bundles will be established by using ANSYS FLUENT software. The transient solidification process was analyzed by examining the key parameters including the transient temperature field, solidification fraction and the first blockage location. The effects of the subcooling temperature, Reynolds number and mushy zone parameters on the solidification process were revealed. The findings play an important role in strengthening the understanding of complex solidification behavior in lead-cooled fast reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111172"},"PeriodicalIF":1.9,"publicationDate":"2024-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163500","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 and Validation of the Monte Carlo-based multi-cycle neutronic Simulations methodology for HFETR","authors":"Yi Xia,&nbsp;Xingjie Peng,&nbsp;Changhu Kang,&nbsp;Liyong Ma,&nbsp;Runqi Liu,&nbsp;Chang Liu,&nbsp;Jiyang Song,&nbsp;Yao Lu,&nbsp;Yuhao He","doi":"10.1016/j.anucene.2024.111154","DOIUrl":"10.1016/j.anucene.2024.111154","url":null,"abstract":"<div><div>The High Flux Engineering Test Reactor (HFETR) is a distinctive reactor that exhibits notable differences in design and operational characteristics compared to commercial pressurized water reactors. The reactor is distinguished by a sophisticated core composition, an adaptable core loading configuration, and pronounced heterogeneity. In order to achieve this, the RMC Monte Carlo transport-burnup coupling code was employed. The developers constructed a sophisticated computational model to simulate global core and localized fuel irradiation. This model effectively captures the physical dynamics of the core across multiple loops, incorporating a range of variables, including fuel burnup, control rod positioning, beryllium poisoning, and particle transport during reactor operations. The methodology comprises comprehensive refueling and fuel management calculations for an array of fuel types and burnable poisons, coupled with the formulation of a multi-cycle neutronic calculation framework. An evaluation of core reactivity and average point burnup for different fuel irradiation cycles reveals a high alignment between the calculated and experimental data, with an average relative error below 1000 pcm and 10%. This outcome provides substantial evidence supporting the accuracy of the HFETR multi-cycle neutronic simulation methodology.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111154"},"PeriodicalIF":1.9,"publicationDate":"2024-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163516","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":"Temperature flattening and thermoelectric conversion modeling of TOPAZ-II thermionic reactor","authors":"Yao Wu,&nbsp;Zhonghe Sun,&nbsp;Liuren Wang,&nbsp;Huanpeng Liu,&nbsp;Guodong Liu","doi":"10.1016/j.anucene.2024.111135","DOIUrl":"10.1016/j.anucene.2024.111135","url":null,"abstract":"<div><div>Focus on temperature non-uniformity in thermionic reactors, a code for temperature flattening is developed by taking the 100 kW TOPAZ-II thermionic reactor as an example. Two optimization schemes, central hole and fuel enrichment, were implemented. Comparison with experiments shows the model can predict temperature distribution and volt–ampere characteristics. After optimization, emitter temperature and current density are more concentrated. The maximum temperatures of central hole and enrichment schemes are reduced to 1905 K and 1940 K from 1990 K. The optimized reactor can operate at higher power density. At 2000 K emitter maximum temperature, the powers of central hole and enrichment schemes increase by 21.95% and 12.59%. A comparative study shows that increasing emitter maximum temperature from 1800 K to 2000 K can increase reactor energy conversion efficiency from 6.83% to 7.91%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111135"},"PeriodicalIF":1.9,"publicationDate":"2024-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143164240","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":"Modification to aerosols dispersion algorithms over two-dimensional hilly terrain based on wind tunnel atmospheric experiments","authors":"Siyuan Wu,&nbsp;Jinpeng He,&nbsp;Weiguo Gu,&nbsp;Deyi Chen,&nbsp;Baojie Nie,&nbsp;Dezhong Wang","doi":"10.1016/j.anucene.2024.111165","DOIUrl":"10.1016/j.anucene.2024.111165","url":null,"abstract":"<div><div>Accurate prediction of aerosol atmospheric dispersion concentration distribution in airborne effluents from nuclear power plants is very important for environmental assessment and nuclear accident emergency response. This study focuses on the plume dispersion centerline trajectory and concentration distribution of aerosols in the atmospheric boundary layer over a two-dimensional hill model under neutral atmospheric conditions. Through laser measurements in active atmospheric dispersion wind tunnel experiments, the flow field around the hill and the aerosol dispersion concentration distribution released at different heights were obtained. According to the streamlines and considering the lifting effect of hill on plumes, a modification algorithm for plume centerline trajectories and concentration distribution in hilly terrain based on Gaussian dispersion model was proposed. The plume centerline trajectory and concentration distribution simulated by the modification algorithm and Plume Path Coefficient Treatment are compared with the experimental results. The results showed that due to the influence of clockwise circulation on the leeside of the hill, the aerosol dispersion trajectory was separated from the shape of the hill. The modification algorithm has better agreement with the experimental results, with NMSE and FB in <span><math><mrow><mn>0</mn><mtext>–</mtext><mn>0</mn><mo>.</mo><mn>025</mn></mrow></math></span>, in the simulation of the plume centerline trajectory. For the simulation of concentration distributions, the modification algorithm slightly underestimates, but is more stable than Plume Path Coefficient Treatment, which has the performance that goes from far overestimation to underestimation along the downwind direction. This study can provide an accurate evaluation of aerosols atmospheric dispersion over hilly terrain.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111165"},"PeriodicalIF":1.9,"publicationDate":"2024-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163623","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":"Loss of coolant accident diagnosis for large pressurized water reactors based on long short-term memory network","authors":"Ze Zhu,&nbsp;Wenlong Liang,&nbsp;Xianlin Tang,&nbsp;Jiawen Li,&nbsp;Pengfei Wang","doi":"10.1016/j.anucene.2024.111167","DOIUrl":"10.1016/j.anucene.2024.111167","url":null,"abstract":"<div><div>Loss of coolant accident (LOCA) is one of the typical accidents in nuclear power plants (NPPs), which is very difficult to diagnosis accurately using traditional equipment alarm mechanisms or expert experience. This paper carries out the LOCA diagnosis study for large pressurized water reactors (PWRs) using long short-term memory (LSTM) networks. Based on a RELAP5-Simulink coupling model, sample datasets of nine types of LOCA for the target PWR were generated. The break locations were selected as the primary-loop hot and cold legs and steam generator heat transfer tubes, with three break sizes set for each break location. Subsequently, an enhanced LSTM network incorporating ReLU and Dropout layers and batch normalization was trained for accident feature extraction. Simulation results show that the developed LSTM model can realize accurate LOCA diagnosis timely, with a diagnostic accuracy of 97.79%. This study can provide methodological support for intelligent accident diagnosis of PWRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111167"},"PeriodicalIF":1.9,"publicationDate":"2024-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163625","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":"Reduced-order modeling of neutron transport separated in axial and radial space by Proper Generalized Decomposition with applications to nuclear reactor physics","authors":"Kurt A. Dominesey ,&nbsp;Wei Ji","doi":"10.1016/j.anucene.2024.111162","DOIUrl":"10.1016/j.anucene.2024.111162","url":null,"abstract":"<div><div>In this article, we demonstrate the novel use of Proper Generalized Decomposition (PGD) to separate the axial and, optionally, polar dimensions of neutron transport. Doing so, the resulting Reduced-Order Models (ROMs) can exploit the fact that nuclear reactors tend to be tall, but geometrically simple, in the axial direction <span><math><mi>z</mi></math></span>, and so the 3D neutron flux distribution often admits a low-rank “2D/1D” approximation. Through PGD, this approximation is computed by alternately solving 2D and 1D sub-models, like in existing 2D/1D models of reactor physics. However, the present methodology is more general in that the decomposition is arbitrary-rank, rather than rank-one, and no simplifying approximations of the transverse leakage are made. To begin, we derive two original models: that of axial PGD — which separates only <span><math><mi>z</mi></math></span> and the sign of the polar angle <span><math><mrow><mi>α</mi><mo>∈</mo><mrow><mo>{</mo><mo>−</mo><mn>1</mn><mo>,</mo><mo>+</mo><mn>1</mn><mo>}</mo></mrow></mrow></math></span> — and axial-polar PGD — which separates both <span><math><mi>z</mi></math></span> and the full polar angle <span><math><mi>μ</mi></math></span> from the radial domain. Additionally, we grant that the energy dependence <span><math><mi>E</mi></math></span> may be ascribed to either radial or axial modes, or both, bringing the total number of candidate 2D/1D ROMs to six. To assess performance, these PGD ROMs are applied to two few-group benchmarks characteristic of Light Water Reactors. Therein, we find both the axial and axial-polar ROMs are convergent and that the latter are often more economical than the former. Ultimately, given the popularity of 2D/1D methods in reactor physics, we expect a PGD ROM which achieves a similar effect, but perhaps with superior accuracy, a quicker runtime, and/or broader applicability, would be eminently useful, especially for full-core problems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"213 ","pages":"Article 111162"},"PeriodicalIF":1.9,"publicationDate":"2024-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143163624","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}