Journal of Nuclear Materials最新文献

{"title":"Role of severe plastic deformation on mechanical behavior of irradiated materials: A case study with Nb-1Zr alloy","authors":"S. Mondal ,&nbsp;M. Sen ,&nbsp;S.K. Makineni ,&nbsp;P. Ghosh ,&nbsp;A. Sarkar ,&nbsp;R. Kapoor ,&nbsp;S. Suwas","doi":"10.1016/j.jnucmat.2024.155487","DOIUrl":"10.1016/j.jnucmat.2024.155487","url":null,"abstract":"<div><div>In this investigation, the effect of 5.6 MeV proton irradiation on the microstructure and mechanical properties of coarse grained (CG) and nanocrystalline (NC) Nb-1wt.%Zr (NZ) has been analysed. Bulk nanocrystalline microstructure was obtained by subjecting the alloy to room temperature high pressure torsion under 6 GPa hydrostatic pressure and 5 rotations. The CG and NC samples were irradiated at doses of 1.9 × 10¹⁷ p/cm² and 1.8 × 10¹⁷ p/cm², respectively. Microstructural parameters like crystallite size, dislocation density, and dislocation arrangements were studied in detail using X-ray line profile analysis (XLPA) by Convolutional Multiple Whole Profile (CMWP) fitting. Microscopic observations were made with electron microscopy techniques in the scanning and transmission modes. Differential Scanning Calorimetry (DSC) was performed to estimate the concentration of vacancies after HPT processing and irradiation. Tensile tests of irradiated CG and NC irradiated samples were performed and compared to those in unirradiated conditions. In the NC condition, not only did the irradiated sample show higher ultimate tensile strength but also twice the amount of uniform elongation as compared to the irradiated CG sample. The fracture surface clearly exhibited this higher plasticity post-irradiation in the NC samples. The change in deformation mechanisms due to nano-structuring of the microstructure has been anticipated to be a reason for the increase in ductility in a single-phase alloy has been explained thereafter.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155487"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652815","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Synthesis and characterization of multi-component borosilicate glass beads for radioactive liquid waste immobilisation","authors":"Sitendu Mandal ,&nbsp;Gattu Suneel ,&nbsp;Jayaprakasam Selvakumar ,&nbsp;Kaushik Biswas ,&nbsp;Srikrishna Manna ,&nbsp;Sourav Nag ,&nbsp;Balram Ambade","doi":"10.1016/j.jnucmat.2024.155485","DOIUrl":"10.1016/j.jnucmat.2024.155485","url":null,"abstract":"<div><div>High-level radioactive liquid waste (HLW) is immobilized in a glass matrix through a process called vitrification. In this process, HLW and glass-forming oxides are combined in a pre-determined ratio within a glass melter to produce a vitrified waste form. The properties of this waste form, including its ability to accommodate different radioactive isotopes, depend on the composition of the base glass.</div><div>In the present study, multi-component amorphous borosilicate-based glasses (SiO₂-B₂O₃-Na₂O-TiO₂-Fe₂O₃-CaO-K₂O) in bead form (diameter 2–3 mm) were developed. The elemental composition of the glass beads (GBs) was analyzed using an optical emission spectrometer. Additionally, the GBs underwent various physico-chemical analyses, including functional group identification, thermal, electrical, and mechanical properties, as well as viscosity and chemical durability assessments, to identify the optimal glass compositions. The influence of Na₂O on the pouring temperature was also examined. Crushing strength and attrition rate measurements were conducted to confirm the suitability of GBs for remote feeding into the melter. The GBs developed in the study are unique, with significant potential for worldwide use in vitrification facilities, particularly in continuous vitrification systems employing Joule Heated Ceramic Melter (JHCM) technology.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155485"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652878","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effects of Ce addition on the morphology, crystal and metal/oxide interface structures of nanoparticles in FeCrAl-ODS steels","authors":"Tian-Xing Yang,&nbsp;Peng Dou,&nbsp;Chang-Jun Zhou","doi":"10.1016/j.jnucmat.2024.155484","DOIUrl":"10.1016/j.jnucmat.2024.155484","url":null,"abstract":"<div><div>FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials in generation IV nuclear reactors due to its exceptional macro-properties. To address the stringent performance requirements of supercritical water-cooled reactors (SCPWRs), two FeCrAl-ODS steels, i.e., 3Al–0.1Ti (Fe–16Cr–3Al–0.1Ti–0.34Y₂O₃) and 2Al–0.1Ti–0.35Ce (Fe–16Cr–2Al–0.1Ti–0.35Ce–0.36Y₂O₃), were developed. This study aims to investigate how Ce addition influences the microstructure and the formation mechanisms of various oxides in ODS steels. Therefore, the grain &amp; nanoparticle morphologies, and crystal &amp; interface structures of nano-scale oxides of the two ODS steels were studied by transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM) and high-resolution transmission electron microscopy (HRTEM). The mean grain diameter of 3Al–0.1Ti and 2Al–0.1Ti–0.35Ce is 1.1 μm and 0.82 μm, respectively. Compared with 3Al–0.1Ti, the average diameter of particles of 2Al–0.1Ti–0.35Ce is relatively smaller. The results indicate that adding Ce can refine the grains and nano-sized particles. For 3Al–0.1Ti, the main particles are Y–Al–O with a proportion of ∼81.4 %. For 2Al–0.1Ti–0.35Ce, the main particles are Y–Ce and Y–Ti oxides with quantity ratios of ∼52.2 % and ∼22.1 %, respectively, while the quantity ratio of Y–Al oxides is only 12.3 %. This indicates that adding Ce can impede the occurrence of Y–Al–O while facilitating the generation of Y–Ce–O. Moreover, it is the first time that Y₂Ce₂O₇ oxide has been detected in yttria-added ODS steels with Ce. The findings obtained from this study provide key insights into the mechanisms of oxide formation &amp; polymorphic transitions, and microstructural differences due to Ce addition. This will provide pivotal direction for the optimization of alloy compositions, promoting the innovation of ODS steels. Additionally, the feasibility analysis of the two ODS steels indicates their applicability to the SCPWR fuel cladding.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155484"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652828","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Molecular dynamics simulation of punched loop detachment during helium bubble growth in nickel","authors":"A-Li Wen ,&nbsp;He-Fei Huang ,&nbsp;Zhen-Bo Zhu ,&nbsp;Wei Zhang ,&nbsp;Fei-Fei Zhang ,&nbsp;Cui-Lan Ren ,&nbsp;Ping Huai","doi":"10.1016/j.jnucmat.2024.155479","DOIUrl":"10.1016/j.jnucmat.2024.155479","url":null,"abstract":"<div><div>The coarsening of helium (He) bubbles in nickel-based alloys significantly impacts their service performance. Understanding the underlying mechanisms is crucial for ensuring the long-term durability and reliability of these alloys in reactor radiation environments. Molecular dynamics simulations of single bubble growth at temperatures of 300 and 900 K were conducted using the sequential He atom injection method to investigate the He bubble growth and evolution in nickel. A noteworthy phenomenon observed during bubble growth is the detachment of punched prismatic loops. The critical bubble size for punched loop detachment can be reduced by growing the bubble at a slower rate or lower temperature. The reduction is attributed to the additional time available for the punched loop to dissociate or the higher pressure within the bubble pushing it out. Meanwhile, the formation mechanism of bubble-loop complexes is explored through the interaction of punched loops with nearby punched loops or bubbles. In addition, the integration of these simulation results with variations in material mechanical performance yields valuable insights for interpreting material degradation. This study provides a foundation for improving in-reactor service performance, contributing to a broader understanding of the complex interplay between helium bubble coarsening and material behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155479"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652879","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"The role of irradiation-enhanced interstitial diffusion in over-pressurizing fission gas bubbles in UO2","authors":"M.W.D. Cooper,&nbsp;C. Matthews,&nbsp;D.A. Andersson","doi":"10.1016/j.jnucmat.2024.155452","DOIUrl":"10.1016/j.jnucmat.2024.155452","url":null,"abstract":"<div><div>Fission gas bubbles in UO₂ nuclear fuel have been observed to exhibit pressures in excess of the equilibrium bubble pressure; however, the cause of bubble over-pressurization has not yet been demonstrated. The mechanical interaction between a bubble and the surrounding matrix or grain boundary depends on the internal pressure of the bubble and local stress state, such that over-pressurized bubbles are thought to be responsible for fragmentation and pulverization, when exposed to a temperature ramp. Here, we investigate the role of U interstitials, produced through irradiation, in over-pressurizing bubbles by using a combined molecular dynamics (MD) and cluster dynamics approach. Firstly, the energies for the capture of interstitials and vacancies by bubbles have been determined from MD as a function of the ratio of gas atoms to vacancies that make up the bubble. Secondly, these reaction energies have been implemented in the cluster dynamics code Centipede to predict bubble over-pressurization as a function of temperature for typical fission rates. It was found that there is a transition from low pressure bubbles (at high temperatures) to high pressure bubbles (at lower temperatures). The cause of this behavior was shown to be the creation of irradiation-induced interstitials that are highly mobile relative to vacancies at low temperature; whereas, vacancies are sufficiently mobile at high temperatures to limit bubble pressures. This result supports the hypothesis that over-pressurized bubbles form during steady-state operation and that this behavior is highly sensitive to the local pellet temperature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155452"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652837","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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 advanced hydride reorientation model for Zircaloy cladding and its experimental validation","authors":"Changhyun Jo,&nbsp;Dahyeon Woo,&nbsp;Youho Lee","doi":"10.1016/j.jnucmat.2024.155445","DOIUrl":"10.1016/j.jnucmat.2024.155445","url":null,"abstract":"<div><div>Hydride reorientation, which occurs under hoop stress during cooling, stands out as a primary mechanism for material degradation in spent fuel management. The radial hydride fraction (RHF) is strongly involved in the mechanical integrity of cladding, highlighting the necessity for a robust modeling framework for quantitative analysis. However, the predictability of previous thermodynamic models for hydride reorientation in reactor-grade Cold Worked Stress Relieved (CWSR) Zircaloy has been hindered due to the intricate nature of hydride reorientation and the difficulties in characterizing microstructures. Recent successful EBSD characterization of reactor-grade CWSR Zircaloy has revealed valuable insights into microstructural characteristics of hydrides, enabling advancements in the modeling framework of hydride reorientation. This study aims to develop a thermodynamic model specifically focused on predicting the RHF. The developed thermodynamic model, based on classical nucleation theory, integrates aforementioned microstructural findings, combined with the Hydride-Nucleation-Growth-Dissolution (HNGD) model to capture transient precipitation behavior during cooling. Extensive experimental validations demonstrate enhanced predictability of the model. Additionally, the study examines the sensitivities of hydride reorientation to hydrogen concentration, applied stress, and cooling rate. It also provides predictions on reorientation behavior for engineering implications such as extension of wet storage, matrix hardening, recrystallization, and thermal cycling, supported by plausible explanations rooted in the underlying physical mechanisms elucidated through the model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155445"},"PeriodicalIF":2.8,"publicationDate":"2024-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661915","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Corrigendum to ‘Investigation of dissolution behavior of SrO in molten LiCl-KCl salts for heat reduction of used nuclear fuel’ [Journal of Nuclear Materials 562 (2022) 153615]","authors":"Dokyu Kang ,&nbsp;Nakkyu Chae ,&nbsp;Wonseok Yang ,&nbsp;Seokjoo Yoon ,&nbsp;Richard I. Foster ,&nbsp;James T.M. Amphlett ,&nbsp;Sang-Eun Bae ,&nbsp;Eun-Young Choi ,&nbsp;Sungyeol Choi","doi":"10.1016/j.jnucmat.2023.154746","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154746","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"588 ","pages":"Article 154746"},"PeriodicalIF":3.1,"publicationDate":"2023-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41086211","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"FeCrAl fuel/clad chemical interaction in light water reactor environments","authors":"Haozheng J. Qu ,&nbsp;Maria Higgins ,&nbsp;Hamdy Abouelella ,&nbsp;Fabiola Cappia ,&nbsp;Jatuporn Burns ,&nbsp;Lingfeng He ,&nbsp;Caleb Massey ,&nbsp;Jason Harp ,&nbsp;Kevin G. Field ,&nbsp;Richard Howard ,&nbsp;Rajnikant V. Umretiya ,&nbsp;Andrew K. Hoffman ,&nbsp;Janelle P. Wharry ,&nbsp;Raul B. Rebak","doi":"10.1016/j.jnucmat.2023.154717","DOIUrl":"10.1016/j.jnucmat.2023.154717","url":null,"abstract":"<div><p><span><span>This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” </span>diffusion multiples of FeCrAl alloys and ceramic UO</span>₂<span><span><span><span> fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent </span>degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) </span>cladding materials<span><span><span> in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of </span>amorphous<span> Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a </span></span>tritium </span></span>permeation<span> barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO</span></span>₂ interface than C35M due to the higher bulk Cr:Al ratio.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154717"},"PeriodicalIF":3.1,"publicationDate":"2023-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45283500","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Preparation of Tc-NpO2 metal-ceramic compositions and their imitators (Re, Th, Nd) for long-term safe storage of long-life fission products","authors":"A.M. Fedoseev ,&nbsp;A.A. Bessonov ,&nbsp;A.V. Sitanskaia ,&nbsp;M.A. Volkov ,&nbsp;A.G. Volkova ,&nbsp;M.N. Sokolova ,&nbsp;D.V. Ryabkov ,&nbsp;K.K. Korchenkin ,&nbsp;K.E. German","doi":"10.1016/j.jnucmat.2023.154711","DOIUrl":"10.1016/j.jnucmat.2023.154711","url":null,"abstract":"<div><p>The preparation routes for ceramic-metal matrices (CMM) Tc-NpO₂ and their imitators Tc(Re)-ThO₂ and Tc(Re)-Nd₂O₃ with different metal-oxide ratios has been investigated and the physical and chemical properties (such as leaching, thermal stability) of the products were determined. The structure of [Np(H₂O)₃(TcO₄)₄]×H₂O has been determined by single crystal X-ray diffraction analysis. Thermal decomposition of neptunium pertechnetates (IV and VI) as well as their imitators neodymium and thorium perrhenates were considered in detail. The surface morphology for CMM Tc-NpO₂ and the leaching rate of each radionuclide was experimentally determined.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154711"},"PeriodicalIF":3.1,"publicationDate":"2023-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41368222","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Calculation of neutron-induced radiation damage to gold and lutetium-aluminium microstructures using MCNP6.2","authors":"S.T. Temaugee ,&nbsp;L. Bedhesi ,&nbsp;R.D. Mavunda ,&nbsp;G.C. Daniels ,&nbsp;S.H. Connell ,&nbsp;I.T. Usman ,&nbsp;E. Chinaka","doi":"10.1016/j.jnucmat.2023.154707","DOIUrl":"10.1016/j.jnucmat.2023.154707","url":null,"abstract":"<div><p>Interaction of particles and photons with materials in extreme radiation environments like the nuclear reactor may lead to basic alterations in the microstructural properties of crystalline solids. These changes accumulate over time into defects on the macrostructure translating to changes in the material's physical and mechanical properties. Studying the level of damage in the materials requires a good prediction of damage using statistical approaches like Monte Carlo simulations. This study aims to calculate the level of damage to Gold (Au) and Lutetium-Aluminium (Lu-Al) due to neutron irradiation, owing to the applications of the materials in reactor technology and other extreme radiation environments. Neutron fluxes, displacement per atom (dpa) rates, and heat deposition expected during irradiation were calculated with Monte Carlo N-particle transport code, MCNP6.2, using the SAFARI-1 research reactor model. The total neutron flux incident on gold and Lutetium-Aluminium was <span><math><mn>2.26</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mn>11</mn></mrow></msup></math></span> n<!--> <!-->cm⁻² <!-->s⁻¹ and <span><math><mn>6.94</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mn>12</mn></mrow></msup></math></span> n<!--> <!-->cm⁻² <!-->s⁻¹ respectively, while the dpa rate in Au and Lu-Al was estimated to be <span><math><mn>1.96</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mo>−</mo><mn>7</mn></mrow></msup></math></span> s⁻¹ and <span><math><mn>6.56</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mo>−</mo><mn>5</mn></mrow></msup></math></span> s⁻¹. The calculated neutron dpa rates and fluence for the materials suggest an elevated level of damage to the microstructure of the materials.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154707"},"PeriodicalIF":3.1,"publicationDate":"2023-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0022311523004750/pdfft?md5=7b0fbb986e07b06a3aa8b0854ddd32a4&pid=1-s2.0-S0022311523004750-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42600575","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}