Journal of Nuclear Materials最新文献

{"title":"Corrigendum to “Identification of Uranium Oxidation States Using Oxygen K-edge Scanning Transmission X-ray Microscopy” [Journal of Nuclear Materials 609 (2025) 155754]","authors":"Rachel E. Lim , Alexander A. Baker , Alexander S. Ditter , S. Olivia Gunther , David K. Shuh , Jack M. Mayer , Matthew A. Marcus , Scott B. Donald , Brandon W. Chung","doi":"10.1016/j.jnucmat.2025.156110","DOIUrl":"10.1016/j.jnucmat.2025.156110","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156110"},"PeriodicalIF":3.2,"publicationDate":"2025-08-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144886039","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":"Characterization and visualization of eutectic melt in B4C powder and pellet-based/304 stainless steel composite control rods under radiative heating","authors":"Zeeshan Ahmed ,&nbsp;Ravinder Kumar ,&nbsp;Avadhesh Kumar Sharma ,&nbsp;Ruicong Xu ,&nbsp;Ryo Yokoyama ,&nbsp;Walter Villanueva ,&nbsp;Hidemasa Yamano ,&nbsp;Marco Pellegrini ,&nbsp;Koji Okamoto","doi":"10.1016/j.jnucmat.2025.156120","DOIUrl":"10.1016/j.jnucmat.2025.156120","url":null,"abstract":"<div><div>The advancement of Sodium-cooled Fast Reactors (Generation IV) faces a key challenge in managing Core Disruptive Accidents (CDAs) caused by eutectic reactions between boron carbide (B₄C) and stainless steel (SS). This reaction causes earlier melting, potentially forming a molten pool containing boron, disrupting neutron balance and causing instability. Therefore, determining boron concentration and distribution in the solidified eutectic pool is critical. This study compares the behavior of B₄C in powder and pellet forms within SS hollow claddings, mimicking control rod designs, to confirm the presence of boron in the eutectic composition and characterize its phases upon solidification. Radiative heating, high-resolution visualization, and quantitative techniques were employed over a temperature of 1372 °C. Findings reveal that the eutectic temperature for pellets exceeds that for powder, with pellets showing fragmentation and powder exhibiting sintering during the failure process. SS cladding formed a melt in both cases; however, in the pellet case, it peeled off, while in the powder case, it ruptured. Visualization methods precisely identified the onset of eutectic melting and associated temperatures, differing between pellets and powder. Characterization techniques, including X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS), and hardness testing, confirmed boron ingress and revealed boride phases such as B_0.9Cr_0.9Fe_1.1, (Cr,Fe)₂B, γ-Fe, (Cr,Fe)₂₃(B,C)₆, and (Cr,Fe)₅B₃ in the eutectic melt. SEM and XPS analyses reveal that both boron and carbon diffuse and precipitate within the eutectic melt, providing new insights into its behavior and relocation dynamics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156120"},"PeriodicalIF":3.2,"publicationDate":"2025-08-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144903273","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":"Combined effects of radiation damage and He accumulation in Y2Ti2O7 borosilicate glass-ceramic composites","authors":"Menghan Jiang ,&nbsp;Anamul Haq Mir ,&nbsp;Mounib Bahri ,&nbsp;Yingjie Zhang ,&nbsp;Keith Arnold ,&nbsp;Nigel Browning ,&nbsp;Karl Whittle ,&nbsp;Maulik Patel","doi":"10.1016/j.jnucmat.2025.156103","DOIUrl":"10.1016/j.jnucmat.2025.156103","url":null,"abstract":"<div><div>Glass-ceramics have been regarded as potential nuclear wasteforms for immobilization of high-level waste. Understanding material modifications due to combined damage by α-decay is the key in predicting long-term evolution of wasteforms. 10 keV in-situ helium implantation experiments were carried out on TEM specimens extracted from Y₂Ti₂O₇ glass-ceramic samples pre-irradiated with 7 MeV Au to understand the kinetics of helium bubble evolution, showing that the accumulation of He can lead to the nucleation of helium bubbles within the amorphous Y₂Ti₂O₇ ceramic phase at a critical concentration of 3.4 at.%. Structural modifications by Au irradiation were characterized using grazing-incidence x-ray diffraction (GIXRD) and scanning transmission electron microscopy (STEM). The STEM results revealed that complete amorphization of the pyrochlore structure occurred at <span><math><mn>5</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mn>14</mn></mrow></msup></math></span> Au/cm². A phase transformation to defect-fluorite structure is also observed at the irradiated interface. Besides, a potential Y-rich phase was identified within the glass matrix, which likely originated from the effects of irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156103"},"PeriodicalIF":3.2,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144892750","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":"Hydrogen trapping and dynamic distribution in iron voids: A molecular dynamics study","authors":"Zi-Qi Li ,&nbsp;Ya-Wen Li ,&nbsp;Yi-Lang Mai ,&nbsp;Wei Wu ,&nbsp;Qiang Qi ,&nbsp;Shouxu Qiao ,&nbsp;Xiao-Chun Li ,&nbsp;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}

{"title":"Effect of aging at 400°C on microstructure and corrosion properties of Fe-13Cr-3.5Al-2Mo-1.5Nb alloy","authors":"Jingyi Tuo ,&nbsp;Shuo Tian ,&nbsp;Xuguang An ,&nbsp;Huabei Peng ,&nbsp;Yiyong Zhang ,&nbsp;Yu Fu ,&nbsp;Qingquan Kong ,&nbsp;Donghai Du ,&nbsp;Xuefei Huang ,&nbsp;Hui Wang","doi":"10.1016/j.jnucmat.2025.156117","DOIUrl":"10.1016/j.jnucmat.2025.156117","url":null,"abstract":"<div><div>The effects of aging time (0–2000 h) at 400 °C on the microstructure and electrochemical corrosion behavior of Fe-13Cr-3.5Al-2Mo-1.5Nb alloy in 1 wt.% NaCl solution were studied. The results showed that after aging at 400 °C, the grains of FeCrAl alloy hardly experienced significant growth, and the precipitates remained fine, uniformly distributed, and nanoscale in size, with more than 95% of the precipitates smaller than 200 nm. Additionally, the self-corrosion potential of the alloy in 1 wt.% NaCl solution shifted positively overall. After aging for 500 and 2000 h, the corrosion current density decreased by 42.9% and 41.9%, respectively, and the corrosion resistance improved. The alloy aged for 2000 h exhibited the higher uniform corrosion resistance and pitting resistance, characterized by a bilayer passivation film, with no niobium oxide phases (NbO, NbO₂, or Nb₂O₅) being detected in the outer oxide layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156117"},"PeriodicalIF":3.2,"publicationDate":"2025-08-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144913626","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":"Oxygen potential and oxygen diffusion data for guiding the manufacture of MOX fuel for fast neutron reactors","authors":"Romain Vauchy ,&nbsp;Yuta Horii ,&nbsp;Shun Hirooka ,&nbsp;Masatoshi Akashi ,&nbsp;Takeo Sunaoshi ,&nbsp;Shinya Nakamichi ,&nbsp;Kosuke Saito","doi":"10.1016/j.jnucmat.2025.156115","DOIUrl":"10.1016/j.jnucmat.2025.156115","url":null,"abstract":"<div><div>Controlling the Oxygen/Metal ratio during the sintering of uranium−plutonium mixed oxide fuels is strategic, especially for fast neutron reactors. Within the frame of understanding the reduction of MOX during its sintering, new oxygen potential data and oxygen chemical diffusion coefficients of U_0.698Pu_0.289Am_0.013O_2−x were determined by thermogravimetry between 1773 and 1923 K on elongated cylindrical dense pellets. An innovative experimental protocol was developed to correlate oxygen chemical diffusion to Oxygen/Metal ratio ranges, and thus to the underlying defect chemistry. Oxygen self-diffusion coefficients were also obtained by combining the oxygen chemical diffusion coefficients with defect chemistry. These new data provide a better understanding of the mechanisms and kinetics of MOX reduction during its manufacturing as a fast neutron reactor fuel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156115"},"PeriodicalIF":3.2,"publicationDate":"2025-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865595","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":"Enhancing the fretting wear resistance of Zr alloy claddings: A CrTiSiN superlattice coating approach","authors":"Jianqiao Yang ,&nbsp;Lifeng Yang ,&nbsp;Yanguang Cui ,&nbsp;Zhuoyu Zhang ,&nbsp;Xunyang Ke ,&nbsp;Fen Zhao ,&nbsp;Xintao Zhang ,&nbsp;Yufan Jiang ,&nbsp;Xianglong Guo ,&nbsp;Junqiang Lu ,&nbsp;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 Ti₂N 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}

{"title":"Chromium doping effects on UO2 grain boundary chemistry: A combined experimental and modeling approach","authors":"Adrien J. Terricabras ,&nbsp;Conor O.T. Galvin ,&nbsp;Maria Kosmidou ,&nbsp;Miguel Pena ,&nbsp;Arjen van Veelen ,&nbsp;William D. Neilson ,&nbsp;Shen J. Dillon ,&nbsp;Michael W.D. Cooper ,&nbsp;David A. Andersson ,&nbsp;Sarah C. Finkeldei ,&nbsp;Joshua T. White","doi":"10.1016/j.jnucmat.2025.156114","DOIUrl":"10.1016/j.jnucmat.2025.156114","url":null,"abstract":"<div><div>Chromium-doped UO₂ 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 UO₂ 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 UO₂, 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}

{"title":"Effects of chromium concentration on the uniform corrosion behavior of FeCrAl alloy in hydrogenated water","authors":"Tenghong Lin ,&nbsp;Zhaolin Shi ,&nbsp;Hui Wang ,&nbsp;Donghai Du","doi":"10.1016/j.jnucmat.2025.156113","DOIUrl":"10.1016/j.jnucmat.2025.156113","url":null,"abstract":"<div><div>FeCrAl alloys are promising accident-tolerant fuel cladding candidates for light water reactors, where balancing irradiation and corrosion resistance is crucial. While lower Cr content generally enhances irradiation hardening resistance, its impact on high-temperature water corrosion remains unclear. Here, we systematically evaluated how Cr concentration (7–13 wt%) affects the uniform corrosion behavior of FeCrAl in simulated PWR environments. The experimental results showed that the metal loss due to corrosion reduces linearly with an increase in Cr concentration. Although 13 wt% Cr yields superior corrosion resistance, reducing Cr content to 7 wt% retains adequate corrosion resistance and demonstrates remarkable long-term stability, as evidenced by oxidation kinetics and nanoscale oxide characterization. These findings reveal that Cr reduction, potentially beneficial for irradiation resistance, does not significantly compromise corrosion performance, enabling cost-effective yet durable cladding designs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156113"},"PeriodicalIF":3.2,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865596","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":"Microstructural evolution and mechanical response of ion-irradiated Fe-9Cr alloys: Insights from nanoindentation","authors":"K. Mulewska ,&nbsp;D. Kalita ,&nbsp;M. Wilczopolska ,&nbsp;W. Chromiński ,&nbsp;P.A. Ferreirós ,&nbsp;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}