Journal of Nuclear Materials最新文献_第6页

Hydrogen trapping and dynamic distribution in iron voids: A molecular dynamics study 铁空隙中的氢俘获和动态分布：分子动力学研究

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-21 DOI: 10.1016/j.jnucmat.2025.156112

Zi-Qi Li , Ya-Wen Li , Yi-Lang Mai , Wei Wu , Qiang Qi , Shouxu Qiao , Xiao-Chun Li , Hai-Shan Zhou

{"title":"Hydrogen trapping and dynamic distribution in iron voids: A molecular dynamics study","authors":"Zi-Qi Li , Ya-Wen Li , Yi-Lang Mai , Wei Wu , Qiang Qi , Shouxu Qiao , Xiao-Chun Li , Hai-Shan Zhou","doi":"10.1016/j.jnucmat.2025.156112","DOIUrl":"10.1016/j.jnucmat.2025.156112","url":null,"abstract":"<div><div>The supersaturated vacancies in the structural materials of nuclear fusion reactor blankets, induced by high-energy neutron irradiation, migrate and aggregate to form voids. Hydrogen (H) isotopes are captured and absorbed by these voids, forming gas bubbles within the voids, leading to H isotope retention and undesirable structural properties. However, most existing studies have primarily focused on small-sized vacancy clusters, with limited attention given to the behavior of H atoms in large-sized nanovoids. This study investigates the dynamic distribution of H in nanovoids of <em>α</em>-iron (Fe). The capture behavior of H atoms by vacancy clusters is calculated using dynamic annealing relaxation and molecular statics methods. Studies indicate that H primarily attaches to the quasi-octahedral interstitial sites at the void boundary in atomic form. Additionally, the number of H atoms absorbed by the vacancy clusters before saturation is linearly correlated with the cluster surface area, while the number of H molecules is linearly proportional to the cluster volume. As the amount of H increases, H molecules are generated in the voids, and the void surface gradually forms saturated H adsorption. After saturation, the H molecules subsequently dissociate into H atoms and diffuse out of the voids. H atoms permeating the Fe lattice displace vacancies and Fe atoms, causing the Fe atoms to collapse inward into the voids. Consequently, voids with a high H-to-vacancy ratio cannot remain stable. This study not only quantifies the capture efficiency and pressure evolution characteristics of nanovoid-H complexes but also provides a theoretical basis for the design of H-resistant alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156112"},"PeriodicalIF":3.2,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902664","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}

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Effect of aging at 400°C on microstructure and corrosion properties of Fe-13Cr-3.5Al-2Mo-1.5Nb alloy 400℃时效对Fe-13Cr-3.5Al-2Mo-1.5Nb合金组织和腐蚀性能的影响

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-20 DOI: 10.1016/j.jnucmat.2025.156117

Jingyi Tuo , Shuo Tian , Xuguang An , Huabei Peng , Yiyong Zhang , Yu Fu , Qingquan Kong , Donghai Du , Xuefei Huang , Hui Wang

引用次数: 0

Oxygen potential and oxygen diffusion data for guiding the manufacture of MOX fuel for fast neutron reactors 指导快中子堆MOX燃料制造的氧势和氧扩散数据

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-16 DOI: 10.1016/j.jnucmat.2025.156115

Romain Vauchy , Yuta Horii , Shun Hirooka , Masatoshi Akashi , Takeo Sunaoshi , Shinya Nakamichi , Kosuke Saito

引用次数: 0

Enhancing the fretting wear resistance of Zr alloy claddings: A CrTiSiN superlattice coating approach 提高Zr合金包层微动磨损性能的CrTiSiN超晶格涂层方法

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-15 DOI: 10.1016/j.jnucmat.2025.156107

Jianqiao Yang , Lifeng Yang , Yanguang Cui , Zhuoyu Zhang , Xunyang Ke , Fen Zhao , Xintao Zhang , Yufan Jiang , Xianglong Guo , Junqiang Lu , Di Yun

{"title":"Enhancing the fretting wear resistance of Zr alloy claddings: A CrTiSiN superlattice coating approach","authors":"Jianqiao Yang , Lifeng Yang , Yanguang Cui , Zhuoyu Zhang , Xunyang Ke , Fen Zhao , Xintao Zhang , Yufan Jiang , Xianglong Guo , Junqiang Lu , Di Yun","doi":"10.1016/j.jnucmat.2025.156107","DOIUrl":"10.1016/j.jnucmat.2025.156107","url":null,"abstract":"<div><div>In this study, the fretting wear behavior of a Zr alloy coated with a CrN/TiSiN multilayer structure was investigated using a lab-scale autoclave fretting apparatus. The multilayer coating exhibited a superlattice architecture consisting of CrN and Ti<sub>2</sub>N units, with each individual layer having a thickness of approximately 50 nm. The CrTiSiN coating demonstrated significantly improved wear resistance compared to the uncoated Zr alloy. The maximum wear loss of the CrTiSiN coating is 36 %, and no visible cracks were observed in the residual CrTiSiN coating. When fretted against a Zirlo dimple, the wear volume and maximum wear depth of the CrTiSiN coated sample were approximately 21 times and 7.1 times lower than those of the uncoated Zirlo specimen. A tribologically induced three-body layer, composed of wear debris containing Zr, Cr, Ti, and O, was observed on the worn surface. The dominant wear mechanism of the CrTiSiN coating was identified as adhesive wear, with a minor contribution from abrasive wear.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156107"},"PeriodicalIF":3.2,"publicationDate":"2025-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144878876","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}

引用次数: 0

Chromium doping effects on UO2 grain boundary chemistry: A combined experimental and modeling approach 铬掺杂对UO2晶界化学的影响：实验与模拟相结合的方法

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-15 DOI: 10.1016/j.jnucmat.2025.156114

Adrien J. Terricabras , Conor O.T. Galvin , Maria Kosmidou , Miguel Pena , Arjen van Veelen , William D. Neilson , Shen J. Dillon , Michael W.D. Cooper , David A. Andersson , Sarah C. Finkeldei , Joshua T. White

{"title":"Chromium doping effects on UO2 grain boundary chemistry: A combined experimental and modeling approach","authors":"Adrien J. Terricabras , Conor O.T. Galvin , Maria Kosmidou , Miguel Pena , Arjen van Veelen , William D. Neilson , Shen J. Dillon , Michael W.D. Cooper , David A. Andersson , Sarah C. Finkeldei , Joshua T. White","doi":"10.1016/j.jnucmat.2025.156114","DOIUrl":"10.1016/j.jnucmat.2025.156114","url":null,"abstract":"<div><div>Chromium-doped UO<sub>2</sub> has been investigated as an Accident Tolerant Fuel (ATF) concept to enhance the performance and safety of Light Water Reactors (LWR). This study explores the impact of varying Cr doping levels on its segregation to and precipitation at grain boundaries of UO<sub>2</sub> through characterization analysis using transmission electron microscopy (TEM) and energy dispersive x-ray spectroscopy (EDS). A broad range of Cr doping levels is examined, from low solubility concentrations (750 ppm) to near-maximum solubility levels (2500 ppm) and extending beyond reported solubility limits (7800 ppm). The study compares these doping levels with undoped UO<sub>2</sub>, evaluating their effects on the atomic concentration of Cr at grain boundaries and grain boundary thickness, all of which are influenced by Cr segregation. Changes in oxidation state were determined via X-ray Absorption Near Edge Structure (XANES). Molecular dynamics simulations are compared to experimental results, discussing concentration evolution, grain boundary type, and segregation energies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156114"},"PeriodicalIF":3.2,"publicationDate":"2025-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144913627","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}

引用次数: 0

Effects of chromium concentration on the uniform corrosion behavior of FeCrAl alloy in hydrogenated water 铬浓度对FeCrAl合金在氢化水中均匀腐蚀行为的影响

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-14 DOI: 10.1016/j.jnucmat.2025.156113

Tenghong Lin , Zhaolin Shi , Hui Wang , Donghai Du

引用次数: 0

Microstructural evolution and mechanical response of ion-irradiated Fe-9Cr alloys: Insights from nanoindentation 离子辐照Fe-9Cr合金的微观组织演变和力学响应：来自纳米压痕的见解

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-14 DOI: 10.1016/j.jnucmat.2025.156109

K. Mulewska , D. Kalita , M. Wilczopolska , W. Chromiński , P.A. Ferreirós , Ł. Kurpaska

{"title":"Microstructural evolution and mechanical response of ion-irradiated Fe-9Cr alloys: Insights from nanoindentation","authors":"K. Mulewska , D. Kalita , M. Wilczopolska , W. Chromiński , P.A. Ferreirós , Ł. Kurpaska","doi":"10.1016/j.jnucmat.2025.156109","DOIUrl":"10.1016/j.jnucmat.2025.156109","url":null,"abstract":"<div><div>Understanding the mechanical behavior of Fe-Cr alloys under irradiation is crucial for their application in nuclear environments. This study investigates the evolution of dislocation structures and their impact on the nanoindentation response of Fe-9Cr alloys in five distinct conditions: (1) non-irradiated pristine material, (2) non-irradiated material annealed at 300 °C, (3) material irradiated with 10 MeV Fe²⁺ ions to 5 dpa at room temperature, (4) material irradiated with 10 MeV Fe²⁺ ions to 1 dpa at 300 °C, and (5) material irradiated with 10 MeV Fe²⁺ ions to 5 dpa at 300 °C. Transmission electron microscopy (TEM) analysis revealed that irradiation at RT leads to a dense distribution of dislocation loops, which act as strong obstacles to dislocation glide, significantly increasing the critical stress required for plastic deformation. In contrast, irradiation at 300 °C results in a lower density of defects in the matrix, with dislocation loops observed near pre-existing dislocation lines. This defect configuration facilitates the formation of dislocation channels, reducing overall obstruction to dislocation motion and leading to a decrease in pop-in stress compared to RT-irradiated samples. However, despite the apparent increase in dislocation mobility, Cr-decorated dislocation loops in the 300 °C-irradiated sample act as pinning sites, impeding the contribution of pre-existing dislocations to plastic deformation and necessitating the nucleation of new dislocations. Recorded mechanical properties, together with microstructural evolution, provide critical insights into the mechanical response of Fe-Cr alloys, offering valuable implications for their performance in nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156109"},"PeriodicalIF":3.2,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144860402","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}

引用次数: 0

Preparation and characterization of AlCrCuFeMoNix high-entropy alloy coatings for accident-tolerance fuel cladding 耐事故燃料包覆用AlCrCuFeMoNix高熵合金涂层的制备与表征

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-13 DOI: 10.1016/j.jnucmat.2025.156104

Julianti Eva , Yayan Zhu , Ruida Jiang , Yanwei Zhang , Rui Lan

{"title":"Preparation and characterization of AlCrCuFeMoNix high-entropy alloy coatings for accident-tolerance fuel cladding","authors":"Julianti Eva , Yayan Zhu , Ruida Jiang , Yanwei Zhang , Rui Lan","doi":"10.1016/j.jnucmat.2025.156104","DOIUrl":"10.1016/j.jnucmat.2025.156104","url":null,"abstract":"<div><div>The investigation of high entropy alloy (HEA) coatings as a promising material for application in nuclear fuel cladding has been motivated due to their exceptional mechanical and functional properties. This study aims to fabricate AlCrCuFeMoNi<sub>x</sub> (x=0.5, 1.2, 1.8 and 2.3) HEA coatings to enhance the accident tolerance of nuclear fuel cladding through magnetron sputtering on zirconium alloy substrates. The mechanical properties of the coatings, as well as their resistance to high temperature steam oxidation and high-pressure pure water corrosion, were investigated. It was found that the coating with x = 0.5 exhibited optimal hardness and Young’s modulus of 8.57 GPa and 170.75 GPa respectively. After subjecting the coated samples to high-temperature and high-pressure pure water corrosion for 3 days, it was observed that aluminum within the coating formed dense spinel NiAl<sub>2</sub>O<sub>4</sub> which effectively inhibits oxygen diffusion, thereby enhancing the corrosion resistance of the coatings. In terms of accident-tolerance performance in reactor environments at temperatures above 1200 °C, it was found that when x = 1.8, dense Al<sub>2</sub>O<sub>3</sub> and oxide featuring a spinel structure generated on the coating surface which significantly reduced the diffusion rate of oxygen ions during high-temperature steam oxidation. The oxidative weight gain of the coated samples was decreased by 70.41 % compared to the uncoated zirconium alloy, demonstrating excellent accident tolerance properties.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156104"},"PeriodicalIF":3.2,"publicationDate":"2025-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865593","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}

引用次数: 0

High temperature chromium coating cracking investigation during tensile tests monitored by acoustic emission 声发射监测拉伸试验中高温铬涂层开裂的研究

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-13 DOI: 10.1016/j.jnucmat.2025.156108

Yanis Taïbi , Ali Charbal , Jean-Christophe Brachet , Elodie Rouesne , Sergio Sao-Joao , Szilvia Kalacska , Morgan Rusinowicz , Guillaume Kermouche

{"title":"High temperature chromium coating cracking investigation during tensile tests monitored by acoustic emission","authors":"Yanis Taïbi , Ali Charbal , Jean-Christophe Brachet , Elodie Rouesne , Sergio Sao-Joao , Szilvia Kalacska , Morgan Rusinowicz , Guillaume Kermouche","doi":"10.1016/j.jnucmat.2025.156108","DOIUrl":"10.1016/j.jnucmat.2025.156108","url":null,"abstract":"<div><div>The present study focuses on in-situ measurements of crack initiation and propagation in first-generation PVD-HiPIMS chromium coatings on M5<sub>Framatome</sub><span><span><sup>1</sup></span></span> cladding substrates using an acoustic emission (AE) device and a tensile test machine. A key novelty of this work is the implementation of a temperature-controlled cracking monitoring system adapted to the cladding geometry under tensile loading. Post-mortem examinations (after different interrupted tensile tests) provide an evaluation of the in-situ method for determining the crack initiation threshold and crack density evolution. The critical strain to crack initiation increases exponentially from 0.4 % at room temperature to 3 % at 350 °C. Above 410 °C, the coating no longer exhibits brittle cracking until reaching high macroscopic imposed strain (up to 30-50 %). Additionally, the crack density decreases more or less linearly with the increasing testing temperature. At higher temperatures, the coating becomes highly ductile, consistently with the increased plasticity of pure chromium. SEM observations of the coating cross-section confirm that cracks do not propagate beyond the coating and that no delamination occurs. Thus, after rapidly reaching crack density saturation, the residual uncracked chromium coating exhibits significant plasticity and widening of the existing Cr cracks while providing slight mechanical reinforcement to the Zr-based cladding up to at least 400 °C.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156108"},"PeriodicalIF":3.2,"publicationDate":"2025-08-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144878878","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}

引用次数: 0

Optical dilatometry and sintering studies of Cr2O3-doped UO2 cr2o3掺杂UO2的光学膨胀测量和烧结研究

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-08-12 DOI: 10.1016/j.jnucmat.2025.156105

Sarah Vallely , Ritesh Mohun , David W. Williams , P. John Thomas , Mattias Puide , David T. Goddard , William E. Lee , Simon C. Middleburgh

{"title":"Optical dilatometry and sintering studies of Cr2O3-doped UO2","authors":"Sarah Vallely , Ritesh Mohun , David W. Williams , P. John Thomas , Mattias Puide , David T. Goddard , William E. Lee , Simon C. Middleburgh","doi":"10.1016/j.jnucmat.2025.156105","DOIUrl":"10.1016/j.jnucmat.2025.156105","url":null,"abstract":"<div><div>Optical dilatometry has been used to study the sintering behaviour of Cr<sub>2</sub>O<sub>3</sub>-doped UO<sub>2</sub> and such observations have been compared to undoped UO<sub>2</sub>, enabling a visual and measurable comparison of their behaviour during this important manufacturing step. Sintering UO<sub>2</sub> and Cr<sub>2</sub>O<sub>3</sub>-doped UO<sub>2</sub> pellets at different temperatures allowed for grain nucleation and growth to be assessed. Optical dilatometry was implemented to visually assess differences in the densification of UO<sub>2</sub> and Cr<sub>2</sub>O<sub>3</sub>-doped UO<sub>2</sub>. The rate of sintering and extent of shrinkage was found to be higher for the doped pellet, with the Cr<sub>2</sub>O<sub>3</sub>-doped pellet decreasing in volume by ∼40 %, compared to only ∼20 % for the undoped pellet. The temperature at which shrinkage commenced was also approximately 100 – 200 °C lower for the doped sample than for the undoped variant. This indicates that there is a difference between the sintering processes of the two systems. The possible contribution of the CrUO<sub>4</sub> secondary phase to large grain growth has also been examined through Differential Scanning Calorimetry (DSC) in reducing conditions. A change occurring at approximately 1250 °C for CrUO<sub>4</sub> was observed. This temperature corresponds to the end of shrinkage and start of grain growth for Cr<sub>2</sub>O<sub>3</sub>-doped UO<sub>2</sub> pellets, as shown in the optical dilatometry experiments, which could be indicative of the commencement of the large-grain growth process.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156105"},"PeriodicalIF":3.2,"publicationDate":"2025-08-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865594","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}

引用次数: 0