Nuclear Engineering and Design最新文献

{"title":"High-temperature oxidation of Zr-4 and Zr-1Nb-O alloys: Influencing factors, oxidation behaviors and mechanisms","authors":"Huanteng Liu ,&nbsp;Donghai Xu ,&nbsp;Guanyu Jiang ,&nbsp;Xueling Fan ,&nbsp;Guangyi Liu","doi":"10.1016/j.nucengdes.2025.113979","DOIUrl":"10.1016/j.nucengdes.2025.113979","url":null,"abstract":"<div><div>Under loss-of-coolant accidents, the oxidation rate of zircaloy rapidly increases, which can cause cladding failure and pose serious safety risks. Thus, enhancing the oxidation resistance of zircaloy is of utmost importance to ensure the safe utilization of a nuclear power. This work provides a comprehensive review on influencing factors and mechanisms for high-temperature oxidation performance of Zr-4 and Zr-1Nb-O alloys. These factors mainly include alloying composition, oxidizing atmosphere, oxidation temperature, pre-oxidation, irradiation, and hydrogen absorption. Oxidation kinetics, behavior, and mechanisms in steam, O₂ and air are thoroughly discussed, and a comparative analysis of oxidation kinetics is presented. Overall, the addition of Nb enhances the oxidation resistance of zircaloy. In air, the oxidation rate of zircaloy is notably faster compared with that in steam and O₂ environments due to the formation of ZrN. At elevated temperatures, the critical size of zirconia increases, leading to a phase transition and a reduction in the volume fraction of monoclinic zirconia. The phase transition makes the zirconia oxide layer crack and less stable. Pre-oxidation at low temperatures in O₂ or steam significantly improves the oxidation resistance of samples. The formation of oxides during the oxidation process of zircaloy is controlled by O^2– diffusion. The breakaway oxidation of zircaloys occurs as a result of the transformation from tetragonal to monoclinic phase of zirconia, as well as stress relaxation of oxides and evolution of oxide morphology when reaching critical thickness.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113979"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621077","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 off-design performance of a hydrogen and electricity joint production system utilizing a very-high-temperature reactor","authors":"Hang Ni ,&nbsp;Xinhe Qu ,&nbsp;Ping Zhang ,&nbsp;Ekaterina Sokolova ,&nbsp;Han Zhang ,&nbsp;Khashayar Sadeghi ,&nbsp;Wei Peng","doi":"10.1016/j.nucengdes.2025.113987","DOIUrl":"10.1016/j.nucengdes.2025.113987","url":null,"abstract":"<div><div>This study introduces a hydrogen and electricity joint production system that utilizes a very-high-temperature reactor and integrates the iodine–sulfur cycle with the steam Rankine cycle. The system can operate under off-design conditions by adjusting the helium mass flow rates in the main and secondary loops, using constant pressure operation (CPO) or sliding pressure operation (SPO) for the power generation loop. The system’s performance is investigated at partial reactor thermal power. The recommended reactor thermal power load ratios range from 63.90 % to 100 % for CPO and from 64.29 % to 100 % for SPO, with the lower limit determined by the steam generator’s hot-end temperature difference. The hydrogen production rate and the system’s electrical power output both decline with a lower load ratio. Within the recommended load ratios, the hydrogen production rates for CPO and SPO range from 129.44 mol/s to 200 mol/s. With a lower load ratio, the power generation efficiency declines, while the hydrogen-electricity efficiency and system’s exergy efficiency first rise and then fall. At a fixed load ratio, the power generation efficiency, hydrogen-electricity efficiency, and system’s exergy efficiency are higher using SPO than those using CPO, indicating better off-design performance using SPO.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113987"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143628098","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 Non-Intrusive Reduced-Order model for rapid analysis of thermal stratification in pressurizer surge line","authors":"Chuzhen Peng,&nbsp;Han Zhang,&nbsp;Yongwang Ding,&nbsp;Lixun Liu,&nbsp;Yingjie Wu,&nbsp;Jiong Guo,&nbsp;Fu Li","doi":"10.1016/j.nucengdes.2025.113996","DOIUrl":"10.1016/j.nucengdes.2025.113996","url":null,"abstract":"<div><div>The pressurizer surge line serves to connect the pressurizer and primary circuit in a PWR (Pressurized Water Reactor) system. However, thermal stratification at its junction can induce distortion and stress, potentially damaging the pipes. Computational Fluid Dynamics (CFD) is a common numerical tool, but its time-intensive nature poses challenges for real-time assessment, especially with multiple parameter variations. To address this issue, we developed a rapid analysis method using a non-intrusive reduced-order model. The experimental design is optimized by incorporating the Generalized Subset Design to minimize sample requirements. The reduced-order model of the temperature field was derived using Proper Orthogonal Decomposition. Off-design cases were predicted using Linear, Radial Basis Function, and Radial Basis Function Neural Network interpolation techniques. The resulting temperature field was utilized for stress analysis in the pipe structure. Results indicate that linear interpolation performs best, with a maximum CvRMSE (Coefficient of Variation of the Root Mean Square Error) of 0.038 for temperature and a maximum RMSE(Root Mean Square Error) of −0.02% in predicting the maximum equivalent stress. The Radial Basis Function interpolation is slightly inferior to linear interpolation. It better fits the thermal stratification region but lacks accuracy in identifying its boundaries. This inaccuracy is more sensitive to equivalent stress, resulting in a maximum stress deviation of −0.08% for sharp boundaries. Additionally, the Radial Basis Function Neural Network is unsuitable for current study due to insufficient sample size, resulting in a maximum stress identification deviation of −3.8%. Finally, the POD coefficient is used as a independent variable to interpolate the maximum Von-Mises stress, and the relative errors were controlled within 5%. This study provides a rapid and accurate method to evaluate the temperature distributions and the maximum stresses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113996"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621075","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":"Investigating the possibility of using iridium as a burnable absorber in new fuel pellet designs of VVER-1200 for reactivity management","authors":"Sultan J. Alsufyani ,&nbsp;Nassar Alnassar ,&nbsp;Mohammed Sallah ,&nbsp;Mohamed A.E. Abdel-Rahman ,&nbsp;Naima Amrani ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2025.113989","DOIUrl":"10.1016/j.nucengdes.2025.113989","url":null,"abstract":"<div><div>Searching for the optimal design of the fuel assembly and the material needed to manage the reactivity in the nuclear reactor is still vital. Therefore, new geometry configurations and materials have been investigated in this work to handle the excess reactivity. Two designs of burnable absorber fuel pellets concentric shell model (CSM) and outer shell model (OSM) have been proposed for VVER-1200 assembly. Four BA materials including Gd₂O₃, Eu₂O₃, Er₂O₃ and Ir₂O₃ have been suggested to be investigated in the proposed fuel pellet designs. Various dimensions and concentrations of the suggested BAs have been studied in the OSM and CSM to obtain the optimum model. From both designs, the optimal models from k_inf behavior point of view have been selected and integral studies have been done on them. The neutronic analysis confirms the effectiveness of using erbium and iridium in the suggested models in managing the excess reactivity of VVER-1200.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113989"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610019","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":"Convolutional neural network use in fast neutron computed tomography for void fraction measurements","authors":"Qasim Siddiq,&nbsp;Garik G. Patterson,&nbsp;Basma Foad,&nbsp;David R. Novog","doi":"10.1016/j.nucengdes.2025.113975","DOIUrl":"10.1016/j.nucengdes.2025.113975","url":null,"abstract":"<div><div>Subchannel analysis codes are used for safety analysis and require suitable data to accurately model complex conditions that occur in fuel bundle geometries. To ensure accurate modelling of the complex two-phase flow phenomena in a reactor bundle, experimental data of the void distribution are required. Computed tomography provides a non-invasive method of measurement of the 2D or 3D distribution of void fraction within a bundle geometry. However, thick pressure vessels in full-scale conditions make it difficult to image bundle geometries and maintain contrast of the internal structures such as water/vapour when using photon-based sources. Fast neutron systems provide good penetration capability while maintaining sensitivity to the water-vapour contrast within the bundle but require long scan times (order of hours) because of the relatively poor detection efficiency and low source strength, which may preclude the application in full-scale thermal–hydraulic testing scenarios. Therefore, to effectively use Fast Neutron Computed Tomography (FNCT) in thermal-hydraulics safety experiments, the large scan times must be reduced, and the increased noise and decreased image fidelity that accompanies this reduction must be addressed.</div><div>This challenge is addressed in this paper using convolutional neural networks (CNNs), a branch of machine learning that excels in image processing tasks. A Synthetic nuclear fuel bundle image reconstructions training dataset, including noise and blurring effects, was generated using a custom MATLAB and Python. The CNN model uses the dataset to create a mapping between these noisy reconstructed images and their respective ground truth image. In this work, the Residual U-Net model significantly improves image reconstructions, leading to more accurate measurements of subchannel void fraction compared to the initial noisy images. The model is able to predict the subchannel void fraction with a mean absolute percentage error (MAPE) of 3.2 % ± 2.4 % on a subchannel basis. Here, MAPE refers to the mean of the absolute differences between the predicted and true void fractions, expressed in percentage points. This means that a void fraction prediction of 50 % with a true value of 53 % results in a 3 % error, not a relative percentage of the void fraction itself. The void fractions were predicted within 8.43 % of the true values for 95 % of the subchannels, demonstrating that the majority of predictions are highly accurate.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113975"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610017","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Transient local effects analysis during typical accidents of a heat pipe cooled reactor","authors":"Guanghui Jiao,&nbsp;Genglei Xia,&nbsp;Tao Zhou,&nbsp;Minjun Peng,&nbsp;Jianjun Wang","doi":"10.1016/j.nucengdes.2025.113988","DOIUrl":"10.1016/j.nucengdes.2025.113988","url":null,"abstract":"<div><div>Heat pipe cooled reactors (HPCRs) utilize high-temperature heat pipes to export fission heat to the energy conversion unit without requiring a coolant loop, simplifying the reactor core design while enhancing safety. Owing to its high safety standards and compact, lightweight design, this technology is the preferred solution for miniature nuclear energy applications in space and marine environments. Analyzing the transient local effects of HPCRs during typical accidents is essential for precise reactor control. However, a fast and practical computational model is still lacking. Therefore, a system code was established to assess the transient accidents between neutronics and thermal hydraulics within an HPCR core. The code includes a three-dimensional neutron physics model, a full-core thermal dynamics model for hexagonal fuel elements (FTDMH), and an improved multi-node heat pipe flow and heat transfer model. Validation was conducted using experimental data and COMSOL Multiphysics simulation results. Heat pipe failure and control drum misoperation accidents were simulated using the developed code, and the local effects of both accidents were analyzed. The results indicate that when a single control drum undergoes an unintended minor outward rotation, the reactor successfully returns to equilibrium through inherent negative feedback mechanisms. Fuel elements near the affected control drum show slight increases in temperature and power generation compared to normal operations, exhibiting a maximum temperature variation of 1.3  K and a peak power discrepancy of 1.02 %. In case of heat pipe failure, the resulting temperature rise in adjacent fuel elements remains below 30  K, and there is no initiation of a chain reaction leading to further heat pipe failures. The code provides a fast and efficient tool for the design and safety analysis of HPCRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113988"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610018","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":"Predictive control of a heat pipe-cooled reactor based on a neural network model","authors":"Xi Bai ,&nbsp;Jiajun Huang ,&nbsp;Zheyu Chen,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2025.113983","DOIUrl":"10.1016/j.nucengdes.2025.113983","url":null,"abstract":"<div><div>The adoption of heat pipes for heat transfer makes the heat pipe-cooled reactor (HPR) with significant time delays, making it difficult for traditional proportional integral derivative control systems to meet the rapid and precise power control demands. Therefore, neural network-based model predictive control is applied for HPR. NUSTER-100 model predictive control system utilizes a feedforward neural network model as the prediction model. By employing numerical optimization algorithms to obtain the optimal control variables, the system effectively overcomes the impact of significant time delays on the control system, achieving rapid and precise regulation. It can effectively suppress the impact of abnormal states on control performance, and ensure that heat pipe-cooled reactor can operate safely and stably under abnormal conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113983"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143620575","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":"Uncertainty in CFD simulation of reactors and approaches to improve the confidence of simulation results","authors":"Xu Lu ,&nbsp;Wengpeng Jia ,&nbsp;Lingyu Dong ,&nbsp;Yang Li ,&nbsp;Dandan Chen ,&nbsp;Genshen Chu","doi":"10.1016/j.nucengdes.2025.113974","DOIUrl":"10.1016/j.nucengdes.2025.113974","url":null,"abstract":"<div><div>Computational Fluid Dynamics (CFD) plays an important role in the design and optimization of reactor thermal-hydraulic systems and serves as an essential component of the ”virtual reactor” projects globally. However, the credibility of CFD simulations is often questioned due to various sources of error and uncertainty, limiting their broader applications in engineering practice. A rigorous process of Verification, Validation, and Uncertainty Quantification (V&amp;V&amp;UQ) is widely acknowledged as essential for assessing the credibility of CFD simulation results. Nevertheless, existing guidelines and standards remain largely theoretical, lacking practical implementation strategies, which prevents a comprehensive understanding of V&amp;V&amp;UQ among researchers. With the enhancement of computational resources, there is potential to improve the credibility of CFD simulation of reactors by reducing the introduction of errors and uncertainties while enhancing V&amp;V&amp;UQ tools and data accessibility. This paper identifies key sources of error and uncertainty across different stages of CFD simulation of reactors from a software developer’s perspective. Furthermore, it explores approaches to improve the credibility of CFD simulation of reactors, emphasizing four critical areas: minimizing model uncertainty, reducing numerical uncertainty, establishing a robust mesh quality evaluation framework, and advancing supportive tools and datasets. By clarifying sources of error and uncertainty and providing guidance for future developments, this paper aims to contribute to the enhanced credibility of CFD simulation of reactors in the future.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113974"},"PeriodicalIF":1.9,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143601087","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":"Investigation of neutronic performance of some accident-tolerant fuels for small modular advanced high-temperature reactors","authors":"Shlash A. Luhaib ,&nbsp;Nassar Alnassar ,&nbsp;Sultan J. Alsufyani ,&nbsp;Ahmed Salah Khalil ,&nbsp;Mohammed Sallah ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2025.113997","DOIUrl":"10.1016/j.nucengdes.2025.113997","url":null,"abstract":"<div><div>The purpose of this study is to examine some accident-tolerant nuclear fuels for small modular advanced high-temperature reactors (SmAHTR) instead of traditional fuel. Various fuels including UCO, UAl, U₃Si₂, (Th, U)CO, (Th, U)Al and (Th, U₃)Si₂ have been investigated as fuels in the TRISO particle of the SmAHTR to investigate the potential advantages of using these fuel types. Neutronic parameters including infinity multiplication factor, fissile concentration, fertile concentration, reactor grade plutonium concentration, fission products concentration, flux per lethargy and radial power distribution have been examined using the cross-section library ENDF/V.II.0. Some safety parameters have been studied to verify the viability of the suggested fuel. The radioactivity of actinides and non-actinides has been tracked with fuel burnup to distinguish their level. This gives a good indication of the precautions required for treating the spent fuel. The neutronic analysis showed the effectiveness of the suggested fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113997"},"PeriodicalIF":1.9,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143601083","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":"Main outcomes of OECD/NEA THAI-2 project on hydrogen risk and source term investigations: Data application for code validation and containment safety assessment","authors":"S. Gupta ,&nbsp;M. Freitag ,&nbsp;Z. Liang ,&nbsp;F. Funke ,&nbsp;G. Langrock ,&nbsp;S. Beck ,&nbsp;H. Nowack ,&nbsp;A. Bentaib ,&nbsp;L. Cantrel ,&nbsp;J. Ishikawa ,&nbsp;S.W. Hong ,&nbsp;P. Kostka ,&nbsp;J. Glover ,&nbsp;C. Linde ,&nbsp;M. Kotouč ,&nbsp;V. Taivassalo","doi":"10.1016/j.nucengdes.2025.113970","DOIUrl":"10.1016/j.nucengdes.2025.113970","url":null,"abstract":"<div><div>During a core-melt accident, apart from hydrogen, radioactive gases and aerosols are released into the containment, the behaviour of which is of significant importance for determining the radiological source term. The investigation of in-containment combustible gases and fission product behaviour was the<!--> <!-->subject of the OECD/NEA THAI-2 project conducted during 2011–2014. The project focused on experiments on hydrogen behaviour i.e., deflagration in the presence of spray and passive autocatalytic recombiners (PARs) performance in O₂-lean atmosphere, on the interaction of molecular iodine with reactive (silver) and non-reactive (tin oxide) aerosol particles to assess the effect on a potential source term, and on the quantification of the release of gaseous iodine from a flashing jet, representing a PWR steam generator tube rupture scenario during reactor shutdown. The project was supported by 11 countries involving safety organizations, regulatory bodies, research laboratories, universities and industries.</div><div>The experimental programme of the OECD/NEA THAI-2 project strongly contributed to the validation and further development of advanced lumped parameter and computational fluid dynamic codes used for reactor applications by e. g. providing experimental data for code benchmark exercises. The present paper summarizes<!--> <!-->the<!--> <!-->key findings of the project and highlights the importance of project results for mitigation of hydrogen risk and source term related issues. Furthermore, the use of project results by the project partners for code validation and reactor analyses towards management and mitigation of a severe accident in light water reactors is discussed with selected examples.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113970"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592560","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}