Nuclear Engineering and Design最新文献

{"title":"Methodology and confirmatory analysis for density wave oscillation in NuScale US460 helical coil steam generators","authors":"P. Yarsky ,&nbsp;J. Thompson ,&nbsp;M. Bernard ,&nbsp;J. Valverde","doi":"10.1016/j.nucengdes.2025.114442","DOIUrl":"10.1016/j.nucengdes.2025.114442","url":null,"abstract":"<div><div>In the current work, a TRAC-RELAP Advanced Computational Engine (TRACE) model was developed to study density wave oscillations (DWOs) in helical coil steam generator (HCSG) coils of the NuScale US460 design. The NuScale US460 design comprises an integral pressurized water reactor vessel that utilizes natural circulation to provide normal core flow. Heat is removed from the reactor pressure vessel by a pair of intertwined HCSGs. Secondary flow boils and becomes superheated inside the coils of the HCSGs. Under certain conditions, the flow in these coils may become unstable due to the density wave instability mechanism. The regulatory purpose is to evaluate the thermal and hydrodynamic loads of postulated instability to determine if these loads pose a challenge to the integrity of the coils. The Evaluation Model Development and Assessment Process (EMDAP) guided the model development effort. A Phenomena Identification and Ranking Table (PIRT) was used to determine knowledge and assessment gaps relative to key phenomena. Incremental TRACE assessment was performed to support the applicability of TRACE to analyze DWOs in helical coils. The authors considered the possibility of DWOs leading to condensation induced water hammer (CIWH) and developed an approach for conservatively estimating CIWH overpressure loads using TRACE. TRACE calculations were performed over the startup range of the NuScale US460 design where unstable DWOs could occur. TRACE predicted that certain operating conditions during startup would lead to unstable DWOs in the HCSG coils. The most severe instabilities were further evaluated to determine the safety significance. The results show that CIWH is unlikely to occur and that thermal fatigue considerations are manageable.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114442"},"PeriodicalIF":2.1,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046864","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":"Assessment of radiation dose and environmental impact from design basis accidents at the BAEC TRIGA research reactor","authors":"Anisur Rahman ,&nbsp;Abdus Sattar Mollah ,&nbsp;Md Abdul Malek Soner","doi":"10.1016/j.nucengdes.2025.114465","DOIUrl":"10.1016/j.nucengdes.2025.114465","url":null,"abstract":"<div><div>It is crucial to investigate the radiation impacts of the nuclear reactor to protect the environment and human health. A computational approach was used to estimate radiation doses and environmental effects from design basis accidents at the BAEC TRIGA Research Reactor (BTRR). ORIGEN-2.2 was used to estimate the reactor’s radionuclide inventory, while HotSpot 3.1.2 modeled radiation dose and radionuclide dispersion. A site-specific climatic scenario was incorporated to estimate individual and effective tissue doses within a 100 km downwind radius of the reactor. Three accident scenarios were analyzed: releases through the emergency ventilation system (scenario 1), malfunctioning emergency ventilation system (scenario 2), and ground-level releases via structural penetrations (scenario 3). Maximum total effective dose equivalent (TEDE) values were 0.37 mSv, 170 mSv, and 8.5 × 10⁵ mSv, respectively, with the most significant impact occurring in scenario 3, affecting an area of 0.20 km² with doses exceeding 100 mSv. Critical organ doses were identified, with the thyroid, skin, red marrow, and lungs receiving significant doses. While scenario 1 poses minimal risk, scenarios 2 and 3 would require evacuation, sheltering, and iodine prophylaxis. These findings will support the TRIGA reactor operator in updating the Safety Analysis Report (SAR) to ensure compliance with regulatory standards.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114465"},"PeriodicalIF":2.1,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046863","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":"Device for steam cladding oxidation testing at TREAT","authors":"Musa A. Moussaoui,&nbsp;Klint S. Anderson,&nbsp;JunSoo Yoo,&nbsp;Nicolas E. Woolstenhulme","doi":"10.1016/j.nucengdes.2025.114441","DOIUrl":"10.1016/j.nucengdes.2025.114441","url":null,"abstract":"<div><div>To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm⁻² s⁻¹ can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114441"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027521","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 generalized coarse mesh finite difference acceleration for 3D method of characteristics in hexagonal fast reactor simulations","authors":"Xiaoxu Geng ,&nbsp;Yun Hu ,&nbsp;Jiayi Yu ,&nbsp;Chong Wang ,&nbsp;Haodong Shan ,&nbsp;Zhaoshun Wang","doi":"10.1016/j.nucengdes.2025.114429","DOIUrl":"10.1016/j.nucengdes.2025.114429","url":null,"abstract":"<div><div>Numerical simulation has become essential in nuclear reactor design and safety verification due to the high cost and complexity of physical experiments. The Method of Characteristics (MOC) provides high-fidelity neutron transport solutions with inherent parallelism and geometric flexibility. However, conventional MOC implementations face challenges in memory usage, computational efficiency, and convergence speed, particularly for full-core simulations of fast reactors with complex hexagonal assemblies. To address these challenges, this work extends the self-developed 3D neutron transport solver ANT-MOC by implementing a hexagonally track generation module that reduces track counts and memory demands while improving accuracy. The developed generalized coarse mesh finite difference (GCMFD) method, compared with CMFD, naturally supports unstructured hexagonal/pentagonal meshes and enables efficient mesh indexing, adjacency management, and equivalent width calculations in such geometries. In addition, an energy group coarsening acceleration framework is introduced to alleviate the computational burden caused by the fine energy discretization typical of fast reactors. These features make the method particularly suitable for fast reactor simulations with complex geometries and wide energy spectra. Validation on the China Experimental Fast Reactor (CEFR) full-core benchmark shows the framework achieves an <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> error of 18.4<!--> <!-->pcm and fission rate and scalar flux errors below 0.3%. The iteration count decreases from 411 to 144, significantly enhancing convergence efficiency. These results demonstrate a high-precision and efficient neutron transport simulation capability for fast reactor cores, with strong potential for engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114429"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046817","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":"State prediction and analysis of 3D upper plenum of lead–bismuth fast reactor based on model order reduction under transient accidents","authors":"Wenshun Duan ,&nbsp;Carolina Introini ,&nbsp;Antonio Cammi ,&nbsp;Kefan Zhang ,&nbsp;Sifan Dong ,&nbsp;Hongli Chen","doi":"10.1016/j.nucengdes.2025.114447","DOIUrl":"10.1016/j.nucengdes.2025.114447","url":null,"abstract":"<div><div>Accurate prediction of three-dimensional (3D) thermal–hydraulic parameter evolution during transients in lead–bismuth fast reactors is important for safety. Although high-fidelity computational fluid dynamic (CFD) models are accurate, they are computationally expensive for real-time use. Model order reduction (MOR) techniques can alleviate this cost while retaining accuracy. In this work, the upper plenum of the lead–bismuth fast reactor NCLFR-Oil is taken as the object of study. Using the proper orthogonal decomposition (POD)-based MOR method and artificial neural networks (ANN), two different 3D transient analysis frameworks are proposed for different data scenarios. 1) A time-series hybrid model (THM) framework designed for time multiple-query tasks, which enables rapid prediction of future three-dimensional physical fields through nonlinear temporal extrapolation of reduced-order modal coefficients. 2) A hybrid data assimilation (HDA) framework aimed at situations with limited sensor data, where the full 3D field distribution is reconstructed using only sparse temperature measurement points by integrating real-time sensor observations with the MOR. The frameworks enhance computational efficiency significantly, with maximum errors around 0.05. Speed-up ratios of 940 and 713 are achieved for THM and HDA frameworks, respectively. Using only three noisy temperature sensors, the HDA framework accurately reconstructs pressure, temperature, and velocity fields, demonstrating robustness and practical applicability. Sensitivity analyses further confirm reliability under varying sensor numbers and noise levels. This work provides an effective tool for real-time monitoring and safety evaluation under accident conditions, offering high practical value.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114447"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027522","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":"Evaluation of the efficiency of heat supply from nuclear power plants taking into account the heat storage properties of heating networks","authors":"Michael Garievsky,&nbsp;Elena Burdenkova","doi":"10.1016/j.nucengdes.2025.114463","DOIUrl":"10.1016/j.nucengdes.2025.114463","url":null,"abstract":"<div><div>The increasing share of nuclear generation in energy systems necessitates enhanced efficiency and flexibility of nuclear power plants (NPPs), within the bounds of engineering constraints. This article explores ways for increasing NPP efficiency through cogeneration and by using the heat storage properties of district heating networks. It is shown that with the development of centralized heat supply based on nuclear power plants, it is possible not only to reduce fossil fuel consumption and greenhouse gas emissions, but also enhance the operational flexibility of nuclear power plants without altering the reactor power output. Based on thermodynamic calculations of thermal circuits of NPP turbines and economic analysis, a comprehensive assessment of the efficiency of heat supply from NPPs using turbines with uncontrolled steam extraction was performed. Special attention is paid to the thermal storage capacity of heating networks, which allows for the accumulation of thermal energy and redistribution of the heat load over a 24-hour period. Calculations show that utilizing the heat storage properties of heating networks for flexible operation of NPPs increases profits by 5.3 %. For various methods of laying heating networks, the maximum distances from the nuclear power plant to the heat energy consumers at which capital investments pay off have been determined.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114463"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046816","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 framework for state assessment and remaining useful life prediction of control rod drive mechanism roller","authors":"Mengqi Huang ,&nbsp;Zhengyu Du ,&nbsp;Ruibo Lu ,&nbsp;Xiaoji Wang ,&nbsp;Changhong Peng","doi":"10.1016/j.nucengdes.2025.114460","DOIUrl":"10.1016/j.nucengdes.2025.114460","url":null,"abstract":"<div><div>The control rod drive mechanism (CRDM) roller is susceptible to degradation, such as wear and fatigue, under environmental and operational stresses, including temperature, humidity, friction, and impact. To enable timely operating state diagnosis, predict future degradation trends, and support operation control and maintenance decisions, this study develops a framework for CRDM roller state assessment and remaining useful life (RUL) prediction. In the state assessment stage, residual distribution analysis combined with adaptive neighbourhood radius clustering is employed to evaluate roller states under limited input data. In the RUL prediction stage, a particle filter-based degradation model is constructed, with parameters estimated via maximum likelihood and multi-objective optimization, and further corrected using Bayesian theory and backward smoothing to enhance prediction accuracy. Validation on bearing datasets achieved a degradation onset estimation deviation of less than three time steps and a RUL prediction error of 8.19%. At the same time, the application to the CRDM roller yielded an error of 8.96%. These results confirm the framework’s capability for accurate state diagnosis and precise RUL prediction using real-time monitoring data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114460"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027616","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":"Finding leaking fuel in the core based on 134Cs and 137Cs activities during spiking events","authors":"I.A. Evdokimov ,&nbsp;D.V. Dmitriev ,&nbsp;A.G. Khromov ,&nbsp;E.Y. Afanasieva ,&nbsp;P.M. Kalinichev ,&nbsp;A.A. Sorokin ,&nbsp;I.O. Goryushin ,&nbsp;A.Y. Burtsev ,&nbsp;S.P. Zolotarev ,&nbsp;S.V. Babkin ,&nbsp;T.Y. Kvichanskaya ,&nbsp;V.V. Atrazhev","doi":"10.1016/j.nucengdes.2025.114439","DOIUrl":"10.1016/j.nucengdes.2025.114439","url":null,"abstract":"<div><div>At present, the ratio of ¹³⁴Cs and ¹³⁷Cs activities during spiking events is widely considered to be the best indicator of fuel burnup in leaking fuel rods. Evaluations are performed by comparison of the ratio between ¹³⁴Cs and ¹³⁷Cs activities in primary coolant with a reference function of fuel burnup. However, there is no universal correlation between the ¹³⁴Cs/¹³⁷Cs activity ratio and fuel burnup. Since ¹³⁴Cs is produced through neutron capture, the ¹³⁴Cs/¹³⁷Cs activity ratio in fuel depends on neutron energy spectrum. The spectrum is sensitive to fuel enrichment and burnup as well as to fission, moderation and absorption characteristics of the surrounding environment. For this reason, the ¹³⁴Cs/¹³⁷Cs activity ratio as a function of burnup differs for different fuel types and changes from cycle to cycle every time when the fuel loading pattern in the core is varied. Using a unified correlation in practice introduces significant errors in the evaluation of leaking fuel burnup. A new approach is developed for the identification of leaking FAs in the core. In this approach, ¹³⁴Cs and ¹³⁷Cs inventory is calculated with low computational costs for each fuel rod in reactor. A software application has been developed to automatically identify fuel rods in the core that provide the closest agreement with the ¹³⁴Cs/¹³⁷Cs activity ratio measured during the spiking event. The developed approach was validated on NPP data for 14 fuel cycles, each containing a single leaking FA. None of the 14 leaking FAs matched the burnup range predicted by the standard technique, with difference ranging from ∼ 10 to 35<!--> <!-->MWd/kgU. In contrast, the new approach accurately identified all 14 leaking FAs and prioritized them for leakage testing during reactor outage.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114439"},"PeriodicalIF":2.1,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020687","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":"Cesium removal from aqueous solutions using mineral adsorbents: Mechanisms, kinetics, and thermodynamics","authors":"Shizong Wang ,&nbsp;Jianlong Wang ,&nbsp;Xuan Guo","doi":"10.1016/j.nucengdes.2025.114461","DOIUrl":"10.1016/j.nucengdes.2025.114461","url":null,"abstract":"<div><div>This study evaluated cesium (Cs) removal by six minerals, with removal efficiencies at pH 7 of 100 % (zeolite), 85.4 % (montmorillonite), 83.8 % (Na-bentonite), 56.2 % (bentonite), 29.5 % (K-feldspar), and 24.3 % (phlogopite). Sips model-derived capacities reached 15.1  mg/g (zeolite), 11.3 mg/g (montmorillonite), 9.67 mg/g (Na-bentonite), 9.02 mg/g (bentonite), 5.07 mg/g (K-feldspar), and 4.70 mg/g (phlogopite). Thermodynamic analysis confirmed spontaneous, endothermic adsorption. Mechanisms analysis revealed that Cs removal occurred mainly via ion exchange or surface coordination, varying by mineral. Zeolite immobilized Cs in its porous structure; montmorillonite and bentonite used interlayer exchange and hydroxyl coordination; phlogopite formed Al-O-Cs complexes via hydroxyl-fluoride substitution; Na-bentonite enabled Cs-Na exchange with octahedral [Al(OH)₆]^3– stabilization; K-feldspar achieved Cs-O-Al surface bonding. Under 100  kGy ⁶⁰Co irradiation, phlogopite, Na-bentonite, and K-feldspar maintained stable Cs adsorption, while zeolite, montmorillonite, and bentonite showed efficiency reductions of ∼ 8–27 %. Among the tested materials, zeolite, montmorillonite, and Na-bentonite are recommended for their high Cs affinity and radiation durability. These findings highlight the importance of balancing adsorption capacity and radiation resistance in selecting optimal minerals for radioactive Cs remediation and long-term environmental protection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114461"},"PeriodicalIF":2.1,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020686","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":"3D modeling of passive safety condensers with CATHARE3 for a high-power pressurized water reactor","authors":"Michel Belliard,&nbsp;Lucie Groussy","doi":"10.1016/j.nucengdes.2025.114391","DOIUrl":"10.1016/j.nucengdes.2025.114391","url":null,"abstract":"<div><div>In the wake of the Fukushima’s nuclear accident, passive safety systems have gained in appeal to improve a nuclear power plant’s resistance to accidents without requiring external power supplies. SAfety COndensers (SACO) are a promising example of these new systems. These are secondary heat exchangers, immersed in a tertiary pool, attached to the steam piping of Steam Generators (SG). In the event of a failure in the normal water supply to the SG, they take the place of emergency pumps, which require a power supply. Their purpose is to extract residual power from the core by condensation of the secondary steam produced at the SG, and return it in liquid form. Then, the secondary liquid inventory is preserved, preventing the SG from drying out.</div><div>For several years now, CEA, in collaboration with EDF, has been involved in modeling SACO of various designs (straight or “C”-shaped vertical tubes, in a calandria or a small or large pool, etc.) for new reactor concepts. In particular, the 3D modeling of immersed exchangers in a pool, using the CATHARE3 code, challenges the conventional 0D/1D modeling and shows the interest of 3D spatial discretization to better take into account the lateral feed of exchangers. As a result, several 3D SACO models, based on different experimental designs and adapted to the reactor power under consideration, are proposed. There are compared with each other on a typical secondary depressurization transient. Also, the CATHARE3 3D modeling is discussed on a typical station black-out transient for a given type of SACO design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114391"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020683","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}