Journal of Nuclear Materials最新文献

{"title":"The influence of RuO2 crystal morphology on the conductivity of glass melts during vitrification process","authors":"Xilei Duan ,&nbsp;Qiang Zhang ,&nbsp;Xueyang Liu ,&nbsp;Zhenghua Qian ,&nbsp;Kui Zhang ,&nbsp;Guanyu Zhu ,&nbsp;Xiaofeng Liu ,&nbsp;Zhanglian Xu ,&nbsp;Yanbo Qiao","doi":"10.1016/j.jnucmat.2025.156068","DOIUrl":"10.1016/j.jnucmat.2025.156068","url":null,"abstract":"<div><div>In this work, we investigate the effects of acicular and granular RuO₂ crystals on the conductivity of nuclear glass melts. Acicular RuO₂ (RuO₂#a) and granular RuO₂ (RuO₂#g) were prepared by molten salt synthesis (MSS). RuO₂#a appears as acicular crystals up to several tens of micrometers in length, while RuO₂#g appears as clusters of nanoparticles. Both crystals have the same elemental compositions and cell structures. RuO₂#a is more effective in forming a continuous conduction network within the nuclear glass, leading to a greater improvement in the conductivity of nuclear waste glass melts. Electrochemical impedance spectroscopy (EIS) analysis shows that the base glass and the low RuO₂ nuclear waste glass behave as ionic conductors, while the high RuO₂ waste glass (x&gt;0.4 vol%) exhibits metal-like conduction below the glass transition temperature (T_g). The critical volume fractions (<em>X_c</em>) calculated using the generalized effective medium (GEM) equation are 0.1 vol% for RuO₂#a and 0.65 vol% for RuO₂#g. While both acicular and granular RuO₂ crystals improve the electrical conductivity of the nuclear glass melts, the RuO₂#a crystals have a significantly greater effect.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156068"},"PeriodicalIF":3.2,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773138","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":"Solubility and dissolution kinetics of uranium trifluoride in 2LiF-BeF2 molten salt","authors":"Zhangqi Li ,&nbsp;Shizhuan Xu ,&nbsp;Peng Wang ,&nbsp;Jinzhao Zou ,&nbsp;Jian Chen ,&nbsp;Changqing Cao ,&nbsp;Jin Lin","doi":"10.1016/j.jnucmat.2025.156060","DOIUrl":"10.1016/j.jnucmat.2025.156060","url":null,"abstract":"<div><div>Although uranium trifluoride (UF₃) holds promise for applications in molten salt reactors, its high-temperature physicochemical properties in molten salt systems have yet to be thoroughly explored. In this study, the solubility and dissolution kinetics of UF₃ in molten 2LiF-BeF₂ (66–34 mol %, FLiBe) eutectic salt were investigated using the isothermal saturation method within the temperature range of 823 K to 973 K. High-purity UF₃ compacts (∼99.70 %) were synthesized via an optimized solid-phase reaction protocol and subsequent compacting. The UF₃-saturated FLiBe molten salts were prepared by immerse dissolution of nickel mesh-wrapped UF₃ compacts and bulk uranium in the molten salt, eliminating the filtration step. Experimental findings showed that the dissolution equilibrium of UF₃ in the FLiBe salt was 120 h. The solubility of UF₃ exhibited a linear increase (R² &gt; 0.99) from 4.16 wt. % to 12.60 wt. % as the elevation of temperature. Crystallographic analysis confirmed that the typical UF₃ phase was the only uranium-bearing phase throughout the dissolution process, while deconvolution X-ray photoelectron spectroscopy (XPS) verified the exclusive presence of the U³⁺ species under all temperature conditions. Notably, the dissolution kinetics conformed to a mass transport control model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156060"},"PeriodicalIF":3.2,"publicationDate":"2025-07-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738472","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":"Modeling of the vacancy diffusion and dislocation motion induced mesoscale-macroscale creep deformations of dense polycrystalline UO2 under irradiation and high temperature conditions","authors":"Zhexiao Xie,&nbsp;Jing Zhang,&nbsp;Xingdi Chen,&nbsp;Guochen Ding,&nbsp;Shurong Ding","doi":"10.1016/j.jnucmat.2025.156055","DOIUrl":"10.1016/j.jnucmat.2025.156055","url":null,"abstract":"<div><div>As a competitive nuclear fuel, UO₂ is widely applicable in various types of nuclear reactors, including both the current and next-generation reactors. Under extreme irradiation conditions, irradiation creep and thermal creep deformations occur within the fuel, significantly influencing the thermo-mechanical behavior evolution of fuel elements and the service safety of nuclear reactors. To predict the macroscale creep deformations and reveal the underlying mechanism for polycrystalline nuclear fuels, the three-dimensional governing equations are established to describe the multi-field coupling behaviors of vacancy diffusion, dislocation motions and the associated mechanical deformations. The corresponding numerical algorithms and codes for the multi-field coupling calculations are developed to simulate the tensile creep tests for polycrystalline UO₂ under different conditions. The good agreement between the predicted macroscale creep rates and various experimental data validates the effectiveness of the developed models and algorithms. The influences of applied stress, temperature, fission rate and grain size on the multi-scale creep behaviors and dominant mechanisms are analyzed. The research results indicate that: (1) the differing stress-related vacancy equilibrium concentrations at various grain-boundary regions lead to the vacancy flux from the tensile boundaries to other boundaries, resulting in the macroscale elongation along the uniaxial creep tension direction; (2) the irradiation-enhanced diffusion coefficient and grain-boundary vacancy equilibrium concentration are responsible for the increased diffusional creep contributions at the temperatures lower than 1200 K, while the fuel fission rates have a minor impact on dislocation creep contributions within the intermediate temperature range from 1000 K to 1400 K; (3) within the low-temperature range below 1000 K, the macroscopic creep deformations of polycrystalline UO₂ are dominantly attributed to irradiation-induced diffusional creep; at temperatures above 1000 K, the creep mechanism becomes more complex and varies with stress, fission rate and grain size; (4) the fuel fission rate affects the transition temperature from irradiation creep to thermal creep; meanwhile, the grain size influences the activation temperature and the stress required to initiate dislocation annihilation-induced creep. This study offers an effective approach to predict the macroscale creep deformations and elucidate the underlying mechanism for nuclear fuels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156055"},"PeriodicalIF":3.2,"publicationDate":"2025-07-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144767088","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":"Aluminium corrosion reactivation in MKPC and Portland-based wasteforms under simulated alkaline repository conditions","authors":"C. Fernández-García,&nbsp;M.C. Alonso","doi":"10.1016/j.jnucmat.2025.156064","DOIUrl":"10.1016/j.jnucmat.2025.156064","url":null,"abstract":"<div><div>The long-term stability of aluminium (Al) in cementitious wasteforms is challenged by matrix alkalinisation resulting from interaction with alkaline waters from engineered barriers in low- to intermediate-level radioactive waste (LILRW) repositories. This study evaluates the corrosion of Al (A1050) and Al-Mg (AA5754) alloys and the physicochemical evolution of magnesium potassium phosphate cement (MKPC) and Portland-based matrices (CEM I and CEM <em>I</em> + 50% silica fume (SF)) under simulated alkaline exposure. Over 250 days, specimens were subjected to synthetic alkaline water (SPAW), while MKPC was also embedded in a CEM I mortar (AM) to simulate repository contact conditions. In MKPC, alkaline plume diffusion from the external source triggered ion exchange with the matrix, leading to alkalinisation (pH &gt;11), reactivation of Al corrosion, and H₂ release to levels comparable to those observed in CEM I. K-struvite dissolution, Ca–P-rich phase formation, and fly ash hydration were also observed, evidencing long-term destabilisation of the MKPC matrix. In contrast, CEM <em>I</em> + 50%SF, despite being exposed to a highly alkaline plume, maintained the pore solution pH at 10.7 over time due to the advanced hydration of SF, effectively limiting Al reactivity and H₂ generation without matrix alteration.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156064"},"PeriodicalIF":3.2,"publicationDate":"2025-07-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750036","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":"Swelling behavior of high density U3Si2 dispersion fuel","authors":"A. Leenaers ,&nbsp;J. Wight ,&nbsp;S. Van den Berghe ,&nbsp;X. Iltis ,&nbsp;H. Palancher ,&nbsp;B. STEPNIK","doi":"10.1016/j.jnucmat.2025.156061","DOIUrl":"10.1016/j.jnucmat.2025.156061","url":null,"abstract":"<div><div>Within the framework of the LEU-FOREvER and EU-QUALIFY projects (H2020 Euratom work program), an irradiation test called High Performance research Reactors Optimized Silicide Irradiation Test (HiPROSIT) has been performed to demonstrate the acceptable in pile behavior of the high density U₃Si₂ fuel system. The HiPROSIT test consists of four (4) fuel plates which were irradiated at the BR2 reactor in Mol, Belgium in representative reactor conditions as a first step towards qualification of the fuel for use in high power research reactors (HPRR), primarily aimed at LEU conversions. Each of the fuel plates was designed to have U₃Si₂ fuel with various loadings at 4.8, 5.3, and 5.6 gU/cm³ and fuel meat thicknesses. After successful irradiation, all plates have achieved a local maximum burnup of ∼80 % U²³⁵ or ∼4.3E+21 fission/cm³. A stable and predictable behavior of the fuel is evidenced in the nondestructive and microstructural examinations. Furthermore, it was found that the as-fabricated porosity accommodates the initial swelling of the fuel, which limits the swelling of the plate.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156061"},"PeriodicalIF":3.2,"publicationDate":"2025-07-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738473","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":"Corrosion resistance and material optimization in supercritical water oxidation for radioactive waste treatment","authors":"Xinyue Huang,&nbsp;Shuzhong Wang,&nbsp;Yanhui Li,&nbsp;Shenghan Sun,&nbsp;Zhaoxia Mi,&nbsp;Limei Xing,&nbsp;Yuanwang Duan","doi":"10.1016/j.jnucmat.2025.156063","DOIUrl":"10.1016/j.jnucmat.2025.156063","url":null,"abstract":"<div><div>This study addresses the severe corrosion of materials during the supercritical water oxidation (SCWO) treatment of tributyl phosphate (TBP), a key radioactive organic solvent. The corrosion behavior of SS316, Incoloy 800, Incoloy 825, and Inconel 625 was evaluated under oxygen-free, oxidizing, and alkaline conditions in subcritical and supercritical water. Results showed that SS316, although cost-effective, undergoes rapid degradation in oxidizing environments, with a corrosion rate of 1.6 mm/a. In contrast, Inconel 625 maintained excellent corrosion resistance at 0.46 mm/a due to the formation of stable NiCr₂O₄ and Cr₂O₃ oxide layers. Notably, this study demonstrates that phosphate anions derived from TBP decomposition can enhance alloy passivation by forming protective phosphate films. Additionally, alkaline modulation using 1 wt.% sodium hydroxide was shown to reduce corrosion rates across all tested alloys significantly. By coupling corrosion environment control with tailored material selection, a corrosion-resistant SCWO reactor was developed, whose innovative structural design also enhances nuclide separation efficiency, offering valuable engineering insights for SCWO systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156063"},"PeriodicalIF":3.2,"publicationDate":"2025-07-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144724323","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":"Structural dynamics of molten NaCl–MgCl2–LaCl3: A proxy for molten chloride fast reactor fuels","authors":"Woei Jer Ng ,&nbsp;Aydar Rakhmatullin ,&nbsp;Kateryna Goloviznina ,&nbsp;Mathieu Salanne ,&nbsp;Catherine Bessada","doi":"10.1016/j.jnucmat.2025.156056","DOIUrl":"10.1016/j.jnucmat.2025.156056","url":null,"abstract":"<div><div>Chloride salts have emerged as the leading candidate fuel for the new generation of molten salt reactors, particularly molten chloride fast reactors (MCFRs). In France, the ARAMIS-A reactor design is exploring the use of NaCl–MgCl₂–PuCl₃–AmCl₃ as fuel for an actinide-burner reactor. However, literature on the structural dynamics of molten chloride fuel salts remains limited, restricting the accurate modelling of fuel performance and behaviour. In this study, we fill this gap by performing in-situ high-temperature nuclear magnetic resonance (HT-NMR) experiments on NaCl–MgCl₂–LaCl₃, as a surrogate to the (Pu,Am)Cl₃ fuel, to investigate the local structural chemistry of molten chloride salts. By complementing our experimental findings with solid-state nuclear magnetic resonance (SS-NMR) measurements, classical molecular dynamics (MD) simulations and density functional theory (DFT) calculations, we elucidate the complex structural interactions in molten chloride systems, such as lanthanide network formation and chlorine bridging, and illustrate how these interactions vary with temperature and fuel composition. Our results establish a clear relationship between NMR chemical shifts and coordination numbers in molten salts, offering critical insights into the local structural environments that influence the behaviour of actinide-based fuels in molten salt reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156056"},"PeriodicalIF":3.2,"publicationDate":"2025-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750034","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":"Prediction of thermodynamic properties and microstructure of UF4 in LiF-BeF2 and LiF-NaF-KF systems through molecular dynamics simulation","authors":"Xiao-Yu Zhang ,&nbsp;Jian-Xing Dai ,&nbsp;Wei Zhang ,&nbsp;A-Li Wen ,&nbsp;Cui-Lan Ren ,&nbsp;Hai-Ying Fu ,&nbsp;He-Fei Huang","doi":"10.1016/j.jnucmat.2025.156054","DOIUrl":"10.1016/j.jnucmat.2025.156054","url":null,"abstract":"<div><div>This study employs classical molecular dynamics (CMD) simulations to predict the thermodynamic and structural properties of UF₄ in LiF-BeF₂ and LiF-NaF-KF molten salts, which are potential fuel carriers and coolants for Molten Salt Reactors (MSRs). We systematically investigate the density, diffusion coefficients, viscosity, and local structures of these systems at varying UF₄ concentrations (1 % to 25 %) and temperatures (1123 K to 1523 K). Our results reveal a strong linear relationship between density and temperature, while diffusion coefficients and viscosity adhere to the Arrhenius equation. Notably, the local structural analysis highlights the formation of U-F-U and Be-F-Be network structures in UF₄-LiF-BeF₂, and Na-F-U and K-F-U networks in UF₄-LiF-NaF-KF, which significantly influence the physical properties. These findings provide critical insights for reactor design, safety analysis, and fuel cycle optimization.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156054"},"PeriodicalIF":3.2,"publicationDate":"2025-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144724322","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 welding time on microstructural evolution and mechanical properties of 316L stainless steel resistance upset welded joints","authors":"Zhen Li ,&nbsp;Bingbing Chen ,&nbsp;Lu Zhao ,&nbsp;Yuanbo Bi ,&nbsp;Qing Guo ,&nbsp;Shuyue Luo ,&nbsp;Zhen Luo","doi":"10.1016/j.jnucmat.2025.156058","DOIUrl":"10.1016/j.jnucmat.2025.156058","url":null,"abstract":"<div><div>In this study, a resistance upset welding method was employed to achieve reliable joining of 316 L nuclear fuel rods. By varying the welding time, the effects on nugget size, microstructural evolution of the weld zone (WZ), and mechanical properties were systematically investigated. The results align with the development trend of advanced nuclear materials, demonstrating that resistance upset welding with adjustable welding time enables reliable joining of nuclear fuel rods. Welding time had no significant effect on the macroscopic morphology of the weld zone; however, with increasing welding time, the upset length, effective weld length, and expelled flash volume all increased. Compared to the base material (BM), the grain size in the WZ was significantly refined. As welding time increased, both the grain size and the fraction of high-angle grain boundaries (HAGBs) increased, with a distinct transition observed when the welding time exceeded 15 ms. In contrast, the dislocation density and texture intensity decreased with longer welding times. These microstructural changes influenced the average microhardness of the WZ, primarily due to variations in grain size and dislocation density. Tensile test results showed that fracture consistently occurred in the BM, indicating that changes in welding parameters did not compromise the maximum load-bearing capacity of the joints. This finding confirms the reliability of the welding technique employed in this study.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156058"},"PeriodicalIF":3.2,"publicationDate":"2025-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144779999","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":"A machine learning approach to quantify degradation of nuclear fuels and the effects of fission products","authors":"Denise A. Lopes,&nbsp;Rinkle Juneja,&nbsp;Alicia M. Raftery,&nbsp;J. Matthew Kurley,&nbsp;William F. Cureton,&nbsp;Andrew T. Nelson","doi":"10.1016/j.jnucmat.2025.156053","DOIUrl":"10.1016/j.jnucmat.2025.156053","url":null,"abstract":"<div><div>Nuclear fuel performance is critically dependent on understanding the evolution of fuel properties under operational conditions, a complex challenge driven by chemical changes and substantial radiation damage during fission. Traditionally, property evolution has been determined via empirical data collected following irradiation. However, these empirical correlations are limited in their applicability beyond the specific conditions in which they were obtained. This study explores a novel approach to address this challenge by applying materials informatics to develop a machine learning random forest (ML-RF) model that captures the effects of fission products on fuel compounds. The model predicts formation enthalpy (ΔH_f) by leveraging extensive quantum materials property data and correlating it with material descriptors such as composition, atomic and site features, and crystal lattice properties. This ML-RF model enables rapid interpolation across the compositional and structural spaces covered by the training data, thus supporting high-throughput screening and energetic ranking of candidate phases. The model demonstrates the ability to predict ΔH_f with a mean absolute error (MAE) of approximately 0.1 to 0.2 eV/atom across a wide range of compounds, including key nuclear fuel systems (U-O, U-N, U-C, U-Si, and U-Mo). For example, it was used to assess shifts in stoichiometry for UO₂ (O/M) and UN (N/M) fuels, revealing their distinct tendencies in chemical potential variation and enabling preliminary convex hull analyses. Furthermore, the model provides insights into how individual fission products affect fuel properties. Results indicate that larger fission products (e.g., Nd, Pu, Ce) have a more pronounced impact on UO₂, while lighter ones (e.g., Zr) strongly influence UN. The model developed in this work can be used to support the Accelerated Fuel Qualification approach by facilitating preliminary evaluations prior to extensive materials modeling and experimentation. To this end, the trained model has been made available to the fuel community to support ongoing fuel development efforts.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156053"},"PeriodicalIF":3.2,"publicationDate":"2025-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144756977","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}