Journal of Nuclear Materials最新文献

{"title":"Analysis of neptunium oxides produced through modified direct denitration","authors":"Kathryn M. Peruski,&nbsp;Connor J. Parker,&nbsp;Samantha K. Cary","doi":"10.1016/j.jnucmat.2023.154704","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154704","url":null,"abstract":"<div><p>Production of neptunium-237 (²³⁷Np) target materials for plutonium-238 (²³⁸Pu) radioisotope thermoelectric generators (RTGs) for deep space exploration requires advanced chemistry and engineering development. Currently, the domestic Pu-238 Supply Program at Oak Ridge National Laboratory produces neptunium dioxide (NpO₂) for target material using a modified direct denitration (MDD) flowsheet. Although the chemistry, reaction mechanisms, and product characteristics of MDD are well understood for uranium, corresponding studies of the neptunium system are still needed to continue optimization of target material properties, production equipment design, and production flowsheets. The objective of this work is to characterize crystalline phases, morphology, surface texture, and particle size of NpO₂ produced via MDD reactions. Solid-phase characterization techniques, including powder X-ray diffraction (pXRD) and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS), were employed to achieve this objective. Subsequent data processing using the Morphological Analysis for Material Attribution (MAMA) software was performed to analyze particle morphology and size. Broadly, the powders were found to contain a mixture of NpO₂ and Np₂O₅ after denitration with a variety of morphologies. After high-firing, the product was found to be NpO₂ with a typical polycrystalline oxide morphology and a grain size ranging from 0.72 to 0.94 µm. These analyses provide knowledge on the reaction pathway for a non-traditional NpO₂ synthesis method and offer additional unique insight into production-scale environments for transuranic materials.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154704"},"PeriodicalIF":3.1,"publicationDate":"2023-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3405672","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":"Simulations of self- and Xe diffusivity in uranium mononitride including chemistry and irradiation effects","authors":"M.W.D. Cooper ,&nbsp;J. Rizk ,&nbsp;C. Matthews ,&nbsp;V. Kocevski ,&nbsp;G.T. Craven ,&nbsp;T. Gibson ,&nbsp;D.A. Andersson","doi":"10.1016/j.jnucmat.2023.154685","DOIUrl":"10.1016/j.jnucmat.2023.154685","url":null,"abstract":"<div><p><span>A combination of density functional theory and empirical potential atomic scale simulations have been used to determine a model for defect stability and mobility in uranium mononitride (UN), as a function of temperature (</span><em>T</em>) and N₂ partial pressure (<span><math><msub><mrow><mi>p</mi></mrow><mrow><msub><mrow><mi>N</mi></mrow><mrow><mn>2</mn></mrow></msub></mrow></msub></math></span>). Using the model, predictions of hypo-stoichiometry under U-rich conditions compare favorably to CALPHAD calculations using the TAF-ID database. Furthermore, our predictions of U and N self-diffusivity are in good agreement with experiments carried out as a function of <em>T</em> at specific partial pressures under thermal equilibrium. The validated atomic scale data have then been implemented within a cluster dynamics method to simulate irradiation-enhanced defect concentrations. All defects and clusters studied have significantly enhanced concentrations, with respect to thermal equilibrium, as <em>T</em><span> is lowered. The irradiation-enhanced Xe diffusivity is compared to post-irradiation annealing and in-pile experiments. The contributions of various defects and clusters to non-stoichiometry, self-diffusivity, and Xe diffusivity are discussed.</span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154685"},"PeriodicalIF":3.1,"publicationDate":"2023-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46679113","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":"Fe-14Ni-14Cr-2.5Al steel showing excellent corrosion-resistance in flowing LBE at 550 °C and high temperature strength","authors":"Lin Shen ,&nbsp;Guoqing Cao ,&nbsp;Dong Lang,&nbsp;Huabei Peng,&nbsp;Yuhua Wen","doi":"10.1016/j.jnucmat.2023.154703","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154703","url":null,"abstract":"<div><p>We investigated the corrosion behavior, microstructural evolutions, and mechanical properties in an alumina forming austenitic (AFA) Fe-14Ni-14Cr-2.5Al steel after exposure in the flowing lead-bismuth eutectic (LBE) at 550 °C for 4008 h and aging in air for 5000 h. The results showed that the AFA steel exhibited not only excellent corrosion-resistance in the flowing LBE but also a strong stability of austenite and high yield strength (589 MPa) at 550 °C. The excellent corrosion-resistance was ascribed to the rapid formation of a thin Al₂O₃ scale. For the first time, the precipitation of γ'-Ni₃Al phase was observed in the AFA steel after long-term aging at 550 °C.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154703"},"PeriodicalIF":3.1,"publicationDate":"2023-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"2825937","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 channeling effect of symmetrical tilt grain boundaries on helium bubbles in tungsten","authors":"Jingxiao Ren ,&nbsp;Hongxian Xie ,&nbsp;Fuxing Yin ,&nbsp;Guanghong Lu","doi":"10.1016/j.jnucmat.2023.154701","DOIUrl":"10.1016/j.jnucmat.2023.154701","url":null,"abstract":"<div><p><span><span>Helium atoms can migrate easily and aggregate at various </span>crystal defects, such as grain boundaries, which promote the nucleation and growth of </span>helium bubbles<span>, ultimately causing severe radiation damage to nuclear materials. Structure characteristics of grain boundaries with low-angle symmetrical tilt of the [100] axle and their channeling effects on helium bubbles in tungsten<span><span><span> were investigated by molecular dynamics simulations. The triaxial variation curves of the helium bubble size in each grain boundary model and the relationship of the helium </span>diffusion coefficient of each grain boundary to their intrinsic crystal structures are analyzed. The results indicate that the helium bubbles grow into one-dimensional nanochannels at the selected grain boundaries, and there is a significant difference in the maximum number of helium bubbles these grain boundaries can accommodate before forming one-dimensional nanochannels. Furthermore, the existence of helium nanochannels along the 〈100〉 </span>edge dislocation<span> line in the low-angle symmetrical tilt grain boundaries is well verified by molecular dynamics simulation of the helium bubble growth and helium diffusion equations at the 〈100〉 edge dislocation line. This suggests that releasing helium outside of the tungsten matrix through the nanochannel structures of low-angle symmetrical tilt grain boundaries may be a potential strategy to address the root causes of bubble nucleation and swelling of tungsten material.</span></span></span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154701"},"PeriodicalIF":3.1,"publicationDate":"2023-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"54585866","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":"Predicting the particle-agglomeration effect on the equivalent mechanical properties of dispersion nuclear fuel by machine learning","authors":"Yingxuan Dong ,&nbsp;Junnan Lv (Conceptualizion) ,&nbsp;Tao Peng ,&nbsp;Hong Zuo ,&nbsp;Qun Li","doi":"10.1016/j.jnucmat.2023.154697","DOIUrl":"10.1016/j.jnucmat.2023.154697","url":null,"abstract":"<div><p>The optimal design of dispersion nuclear fuel necessitates precise correlations between the particle distribution and mechanical properties<span>, which can be established by combining the numerical method<span><span>, data-driven technique, and machine learning model. In this study, we developed an automated high-throughput workflow to rapidly generate massive number of dispersion nuclear fuel meat models with different particle distributions, and further built the machine learning database for predicting the crucial mechanical properties. Effects of the particle-agglomeration behavior on mechanical properties of the dispersion fuel meat were analyzed in length. The automated workflow includes the entire parametric modeling, parallel computation and post-processing of simulated results, through which we can calculate the equivalent </span>elastic modulus and the maximum Mises stress of the dispersed microstructure with randomly distributed fuel particles. The Fourier distribution function is utilized to characterize the random particle distribution configuration. The analysis suggests that the particle distribution configuration, the volume fraction and the number of agglomerate particles are the crucial factors determining the performance of dispersion nuclear fuel. Furthermore, through utilizing the database constructed by high-throughput computing, the Gaussian process regression algorithm was successfully applied to accurately forecast the mechanical properties in the dispersion fuel meat. This work lays a foundation for optimizing the design of high-performance dispersion nuclear fuel, quickly estimating mechanical properties of composite structures containing randomly dispersed particles, and further extending to analogous systems.</span></span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154697"},"PeriodicalIF":3.1,"publicationDate":"2023-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48597045","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 oxidation model for the prediction of the corrosion risk induced by deposits on cladding surface in PWRs","authors":"Yuan Shen ,&nbsp;Shiyu Tan ,&nbsp;Yunchen Lai ,&nbsp;Hongguo Hou ,&nbsp;Nan Chao ,&nbsp;Caishan Jiao ,&nbsp;Yu Zhou ,&nbsp;Yang Gao","doi":"10.1016/j.jnucmat.2023.154698","DOIUrl":"10.1016/j.jnucmat.2023.154698","url":null,"abstract":"<div><p>The enhancement of the temperature and the concentration of the species on the cladding surface induced by deposits (CRUD) increases the risk of the cladding corrosion in PWR. An oxidation model is developed to accurately predict the corrosion of the Zr-4 cladding in PWRs under the wide range of the boron and lithium concentrations here. Combining this oxidation model with the local thermal-hydraulic model, the CRUD induced local corrosion (CILC) is well predicted and analyzed under various operational conditions, hydrochemical conditions, and CRUD structural parameters. The effects of different conditions on CILC are clearly reflected by the corrosion risk values.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154698"},"PeriodicalIF":3.1,"publicationDate":"2023-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42287478","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":"Microstructure and phase investigation of FeCrAl-Y2O3 ODS steels with different Ti and V contents","authors":"Jaroslaw J. Jasinski ,&nbsp;Tomasz Stasiak ,&nbsp;Wojciech Chmurzynski ,&nbsp;Lukasz Kurpaska ,&nbsp;Marcin Chmielewski ,&nbsp;Malgorzata Frelek-Kozak ,&nbsp;Magdalena Wilczopolska ,&nbsp;Katarzyna Mulewska ,&nbsp;Maciej Zielinski ,&nbsp;Marcin Kowal ,&nbsp;Ryszard Diduszko ,&nbsp;Witold Chrominski ,&nbsp;Jacek Jagielski","doi":"10.1016/j.jnucmat.2023.154700","DOIUrl":"10.1016/j.jnucmat.2023.154700","url":null,"abstract":"<div><p>FeCrAl-based steels are considered promising materials for high-temperature nuclear applications. Over the past years, various compositions have been studied to assess their mechanical properties, structural integrity, and radiation damage resistance. However, the microstructure and phase composition of FeCrAl-ODS steels with the addition of different alloying elements are less commonly studied than pure FeCrAl alloys. The paper presents a novel research path for developing FeCrAl matrix ODS steels with Y₂O₃, Ti, and V additions. The materials synthesis consisted of mechanical alloying of pure metallic components with yttrium oxide in a planetary ball mill under an argon atmosphere. Titanium was added in the amount of 1.0 wt.% to both samples, while 0.5 wt.% of vanadium was added to one sample to verify its impact on the structural stability and hardness. The spark plasma sintering (SPS) technique was used to consolidate the powders. Afterward, the microstructure, chemical composition, phase composition, and hardness were assessed using SEM-EDS, EBSD, TEM-EDS, XRD, XRF, Nanoindentation, and Vickers microhardness. The experimental data reveal rather homogeneous powders after mechanical alloying and dense bulk samples after SPS. The microstructure observations show oxide particles and carbides on the grain boundaries and inside grains of the bcc matrix, which suggests elevated radiation damage resistance. The presence of nanoscale oxide particles (15–50 nm) in the matrix could significantly reduce the impact of aging embrittlement at high temperatures by affecting the chromium diffusion pathways in the ODS steel matrix. The addition of vanadium leads to an improvement of hardness to 4.83±0.43 GPa compared to 3.78±0.34 GPa for the sample without vanadium. Presented experimental results are promising in terms of research and development of FeCr and Al-based ODS materials tailored to operate under harsh conditions in generation IV fission reactors and fusion reactors.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154700"},"PeriodicalIF":3.1,"publicationDate":"2023-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0022311523004683/pdfft?md5=546c45a51975e430176b389606133501&pid=1-s2.0-S0022311523004683-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42448848","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}

{"title":"Irradiation performance and first examinations of Americium bearing blanket fuel from the MARINE irradiation experiment","authors":"S. van Til ,&nbsp;P.R. Hania ,&nbsp;A.V. Fedorov ,&nbsp;E. D'Agata ,&nbsp;D. Freis ,&nbsp;S. Bejaoui ,&nbsp;F. Delage ,&nbsp;A. Gallais-During","doi":"10.1016/j.jnucmat.2023.154699","DOIUrl":"10.1016/j.jnucmat.2023.154699","url":null,"abstract":"<div><p><span>Americium<span> (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like </span></span>²⁴¹Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA.</p><p><span>The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus </span>pellet fuel<span>, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling.</span></p><p>The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154699"},"PeriodicalIF":3.1,"publicationDate":"2023-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41820755","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":"Surface growth, inner defects and interface mixing of Cr coating on Zr alloy irradiated by 5 MeV protons at 400°C","authors":"Huan Chen ,&nbsp;Changyong Zhan ,&nbsp;Wenjuan Gong ,&nbsp;Peinan Du ,&nbsp;Ruiqian Zhang ,&nbsp;Jijun Yang ,&nbsp;Yu Wang ,&nbsp;Tianguo Wei ,&nbsp;Hongyan Yang ,&nbsp;Yu Zou ,&nbsp;Baoqin Fu","doi":"10.1016/j.jnucmat.2023.154696","DOIUrl":"10.1016/j.jnucmat.2023.154696","url":null,"abstract":"<div><p><span><span>Surface, inner and interface damages of Cr coatings irradiated by 5 MeV protons were characterized by XRD<span>, SEM, TEM and </span></span>nanoindentation in this work. After irradiation, a Cr</span>₇C₃<span> growth layer with refined grains was formed on the surface due to the separation and redeposition of diffused Cr atoms. Irradiation-induced defects in the Cr coatings were voids, dislocations, and bubble-vacancies complexes. Importantly, irradiation softening was found for the arc-ion plated Cr coating, which should be due to the deformation recovery and recombination of inherent sinks and irradiation defects. At the coating-substrate interface, the ZrCr</span>₂<span> C14 phase, a multilayered interface and the preferential diffusion<span> were found, meaning chained or layered Cr atoms may dominate the interface diffusion. In mechanism, formation mechanisms of irradiation defects are analyzed and the irradiation-enhanced diffusion, ballistic collision and the sharpening effect are employed to explain the interface mixing. A correlation coefficient of the anisotropic diffusion is proposed in the Cr flux equation for explaining the (110)-dominated diffusion. The results show that irradiation effects of coatings depend on the inherent defects and preferred orientation produced in the preparation process of coatings.</span></span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154696"},"PeriodicalIF":3.1,"publicationDate":"2023-08-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41692729","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":"Self–ion irradiation of high purity iron: Unveiling plasticity mechanisms through nanoindentation experiments and large-scale atomistic simulations","authors":"K. Mulewska ,&nbsp;F.J. Dominguez-Gutierrez ,&nbsp;D. Kalita ,&nbsp;J. Byggmästar ,&nbsp;G.Y. Wei ,&nbsp;W. Chromiński ,&nbsp;S. Papanikolaou ,&nbsp;M.J. Alava ,&nbsp;Ł. Kurpaska ,&nbsp;J. Jagielski","doi":"10.1016/j.jnucmat.2023.154690","DOIUrl":"10.1016/j.jnucmat.2023.154690","url":null,"abstract":"<div><p><span>Ion irradiation<span> may enhance material hardness through crystal defect nucleation and reorganization. In this study, we examine the nanomechanical behavior of high-purity iron samples, comparing the response of pristine specimen to those that have been self–irradiated with 5 MeV ions at 300</span></span>^∘<span><span><span>C. We utilize spherical nanoindentation<span><span> to investigate the nanomechanical response, and we focus on the comprehensive modeling of the self–irradiation effects in high-purity iron through large-scale molecular simulations. Transmission electron microscopy is used in the </span>irradiated regions, at various depths below the nanoindentation imprint, to analyze the nucleation of dislocation networks and the </span></span>plastic deformation<span><span> mechanisms at room temperature. Large scale novel molecular dynamics simulations are conducted to simulate overlapping </span>collision cascades<span><span><span> reaching an irradiation dose with defect density similar to experiments, followed by nanoindentation simulations that display qualitative agreement to experiments. We find that irradiated sample requires higher critical load for the transition from elastic to plastic deformation due to interaction of </span>dislocation lines with the </span>dislocation loops and </span></span></span>point defects formed during the irradiation, leading to hardening.</span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154690"},"PeriodicalIF":3.1,"publicationDate":"2023-08-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43710145","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}