Journal of Nuclear Materials最新文献

{"title":"Sintering characteristics of ThO2 with different powder morphologies","authors":"L. Gubbels ,&nbsp;J.P. Ramos ,&nbsp;J. Vleugels ,&nbsp;M. Verwerft ,&nbsp;L. Popescu ,&nbsp;B. Acevedo","doi":"10.1016/j.jnucmat.2025.156415","DOIUrl":"10.1016/j.jnucmat.2025.156415","url":null,"abstract":"<div><div>In this study, both heterogeneous and homogeneous oxalate precipitation method, as well as hydroxide-carbonate precipitation, were employed to produce ThO₂ powders with distinct morphologies- platelets, cubes, and nanoparticles. These powders were compacted and sintered at 2023 K (1750 °C) under an oxygen potential of -420 kJ mol^-1 for 1 h and 10 h. This approach enabled a systematic investigation of the influence of morphology on packing behavior, densification kinetics, and the resulting microstructure, which were characterized using XRD, SEM, BET surface area analysis, He pycnometry, and immersion density measurements. All powders achieved high final densities (&gt;95 % of the theoretical density) after 10 h of sintering, underscoring their potential for nuclear fuel–related applications. For porous ThO₂ materials intended for low-temperature ISOL target applications, the oxalate-derived powders, particularly those precipitated at room temperature, provided a promising balance between low sinterability and process scalability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156415"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024747","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":"Low-temperature fabrication of Li4SiO4 ceramic pebbles with excellent tritium release performance using SiC as silicon source","authors":"Guangfan Tan ,&nbsp;Xudong Mao ,&nbsp;Yasuhisa Oya ,&nbsp;Yingchun Zhang ,&nbsp;Chang-An Wang ,&nbsp;Yanhao Dong","doi":"10.1016/j.jnucmat.2026.156471","DOIUrl":"10.1016/j.jnucmat.2026.156471","url":null,"abstract":"<div><div>Lithium orthosilicate (Li₄SiO₄) has long been recognized as a promising tritium breeder material due to its high lithium content, excellent tritium release performance and irradiation stability. However, conventional methods of preparing Li₄SiO₄ ceramic pebbles typically require high sintering temperatures, which not only result in lithium loss but also deteriorate the tritium recovery performance. To address the above issues, in this work SiC was used as the silicon source to prepare Li₄SiO₄ ceramic pebbles at low temperature. The experimental results reveal that the phase formation temperature of the Li₄SiO₄ could be reduced to 500 °C. When sintered at 850 °C, Li₄SiO₄ ceramic pebbles have high relative density of 83.6 %, crushing load of 46.8 N, and porosity of 10.6 %, which are superior to those made by the traditional methods. To further evaluate the tritium release performance, the samples were subjected to neutron irradiation and thermal desorption tests. Compared with Li₄SiO₄ prepared via the centrifugal granulation method, the samples exhibited rapid tritium desorption rate of 1.1 × 10⁵ Bq⋅g⁻¹⋅s⁻¹ due to the presence of abundant tritium diffusion channels Three release peaks are observed, at 718 K, 698 K and 670 K, with HTO being the main release form. Isothermal heating experiments of the two samples confirm that the release of tritium in the main form of HTO is governed by diffusion process, while the overall tritium release is also influenced by defect trapping and de-trapping mechanisms. To conclude, the Li₄SiO₄ obtained in this study exhibits excellent tritium recovery performance, which will help to regulate the microstructure of the ceramic breeders and enhance the tritium self-sustaining efficiency in the future.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156471"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024666","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":"Elucidating plastic deformation mechanisms in γ-phase U-Mo-Zr ternary alloys using a machine-learning moment tensor potential","authors":"Hongcai Liang ,&nbsp;Yi Wang ,&nbsp;Wensheng Lai","doi":"10.1016/j.jnucmat.2026.156462","DOIUrl":"10.1016/j.jnucmat.2026.156462","url":null,"abstract":"<div><div>A machine learning interatomic potential based on the moment tensor framework (moment tensor potential, MTP) for the uranium-molybdenum-zirconium ternary system is developed to investigate the plastic deformation mechanisms of U-Mo-Zr alloys. The MTP was trained using an extensive reference database generated from density functional theory calculations, and the resulting potential accurately reproduces the fundamental properties of the <em>γ</em>-phase UMo alloy. Molecular dynamics simulations of the Peierls-Nabarro stresses, modified by the generalized stacking fault energy, reveal that the {110}⟨111⟩ slip system is preferred over the {112}⟨111⟩ and {123}⟨111⟩ systems, and this preference weakens with increasing Zr content. Under uniaxial stretching along the [110] direction, the activation of the {110}⟨111⟩ slip system and the formation of corresponding slip bands are observed. Calculations of the generalized planar fault energy suggest that twinning is more favorable than slip in the {112}⟨111⟩ system, and such priority is enhanced with increasing Zr content. Accordingly, twinning rather than slip is predicted to occur during plastic deformation under certain conditions, which is confirmed by the activation of the {112}⟨111⟩ twinning system during uniaxial stretching along the [100] direction. These findings provide theoretical insights that may promote the processing and application of uranium-based metallic nuclear fuels in future advanced reactor systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156462"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024750","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":"Analysis of attenuation data from the decommissioned ZIon unit 1 reactor pressure vessel beltline weld","authors":"Mikhail A. Sokolov ,&nbsp;Roger E. Stoller","doi":"10.1016/j.jnucmat.2026.156445","DOIUrl":"10.1016/j.jnucmat.2026.156445","url":null,"abstract":"<div><div>In order to examine the attenuation of radiation damage through the thickness of an irradiated reactor pressure vessel (RPV), four segments were acquired from the Zion Unit 1 power plant RPV after the plant was decommissioned. The Zion Unit 1 RPV Beltline Weld Segment 1 was cut into seven blocks, consisting of five base metal and two beltline welds from the high fluence region of the segment. Through-wall test specimens were machined and tested. Specimens included those used for Charpy impact, Master Curve fracture toughness testing, and chemical analysis. The observed through-thickness ductile-to-brittle transition temperatures in the beltline weld deviated significantly from the expected behavior based on the attenuation of fast fluence as a function of depth into the RPV. Beginning at the inside surface, the 41-J Charpy transition temperature was either flat or slightly increasing until the ¾ -T location. The results of a simple, model-based analysis of the Zion beltline weld material that included the irradiation conditions and material chemistry were generally consistent with industry trend curves and the standard attenuation model, rather than the observed data. Although there was no archive material from the RPV available to permit measurement of the unirradiated properties, fracture toughness specimens fabricated from archive surveillance weld were used to obtain an estimate of the initial through-thickness values of the Charpy transition temperature. The Charpy shifts obtained using this approach were similarly in disagreement with the predictions of the US NRC Regulatory Guide 1.99, Rev. 2. However, testing of irradiated Charpy specimens taken from the RPV following post-irradiation annealing (10 hr. at 500 °C) provided a quite different estimate of the unirradiated properties which improved the agreement between the inferred through-thickness Charpy shifts and exponential attenuation model included in Regulatory Guide 1.99/2. The analysis of the Zion data and data obtained in previous post-mortem examinations of decommissioned RPVs indicates that more work is needed to understand the through-thickness properties of RPV materials in order to properly assess through-wall damage attenuation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156445"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024753","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":"Mechanical properties and microstructure of CHSN01 conductor jacket under long-term cryogenic service in future fusion reactors","authors":"Weijun Wang ,&nbsp;Xinlan Ye ,&nbsp;Baozhu Zhang ,&nbsp;Yifei Wu ,&nbsp;Huihui Zhi ,&nbsp;Zhengping Tu ,&nbsp;Xiaogang Shen ,&nbsp;Jinggang Qin ,&nbsp;Jiangang Li","doi":"10.1016/j.jnucmat.2026.156463","DOIUrl":"10.1016/j.jnucmat.2026.156463","url":null,"abstract":"<div><div>The CHSN01 (formerly N50H) austenitic stainless-steel conductor jacket developed for China's future fusion reactor exhibits excellent strength and toughness with a yield strength exceeding 1500 MPa at -269°C, approximately 40% higher than that of the 316LN jacket used in ITER. However, the jacket material must also provide excellent safety performance under ongoing exposure to liquid He temperatures, but the effects of long-term cryogenic exposure on the microstructure and properties of CHSN01 remain unknown. Therefore, this study experimentally simulated the actual preparation and service conditions of the CHSN01 jacket to support its application in fusion reactors. First, the effects of different durations of immersion in liquid He dewar on the mechanical performance and microstructure of the material were characterized by tensile testing, scanning electron microscopy, transmission electron microscopy, X-ray diffraction, and electron backscatter diffraction. Next, the evolution of the material microstructure was studied under various strain conditions to reveal the resulting strengthening and toughening mechanisms. Finally, the feasibility of employing CHSN01 as a cryogenically resistant structural material capable of withstanding the high magnetic fields in next-generation fusion reactors was confirmed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156463"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146170150","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":"Exploring iodide-mediated corrosion of NiCr20 alloys","authors":"Insung Han ,&nbsp;Alejandro Ramos Ballesteros ,&nbsp;Kim Kisslinger ,&nbsp;Simerjeet K. Gill","doi":"10.1016/j.jnucmat.2025.156402","DOIUrl":"10.1016/j.jnucmat.2025.156402","url":null,"abstract":"<div><div>A combined approach of microstructural characterization and thermodynamic calculations was employed to investigate the corrosion at alloy-salt interface of model Ni-based alloy (NiCr₂₀) and molten iodide salt system (LiI-KI). In iodide salt system, the corrosion of the NiCr₂₀ alloy is primarily driven by the selective dissolution of Cr, facilitated by grain boundary attack, which results in intergranular corrosion and the subsequent formation of pore-salt network. Additionally, the preferential formation of a Cr oxide surface film at the alloy-salt interface and salt filled void formation in the NiCr₂₀ alloy in molten LiI-KI salt system is examined and discussed. Understanding the corrosivity of iodide salts that can potentially exist as fission products or serve as surrogate molten salts is critical for assessing material degradation in molten salt reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156402"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922722","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":"Investigations of size effect and normalization models for tensile deformation of A508-III steel miniature specimens","authors":"Cong Chen ,&nbsp;Qi Zhang ,&nbsp;Shengmin Xin ,&nbsp;Wanhuan Yang ,&nbsp;Junwan Li ,&nbsp;Weihua Zhong ,&nbsp;Guangsheng Ning","doi":"10.1016/j.jnucmat.2026.156446","DOIUrl":"10.1016/j.jnucmat.2026.156446","url":null,"abstract":"<div><div>The miniature specimens with different grain sizes and thicknesses were prepared by changing the heat treatment temperature of A508-III steel. The tensile test of miniature specimens at room temperature was carried out, and the mechanism of tensile size effect was analyzed. The results indicate that variations in thickness and grain size lead to a significant size effect in tensile properties. By introducing an influence parameter <em>λ</em> comprehensively characterizing the grain and characteristic size effects of miniature specimen, the Swift mechanical constitutive model considering the size effect is established. To improve the damage model for miniature specimens, the ductile damage evolution parameter of miniature specimen is determined by the finite element aided testing (FAT) method. Based on the established mechanical constitutive model and ductile damage model, the load-displacement curve of the miniature specimen obtained through finite element analysis has an average error of less than 5% compared to the experimental test results. The tensile mechanical properties of miniature specimens are predicted by finite element method, and the normalized models of yield and tensile strength are constructed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156446"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957739","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 Al content and a Cr-N diffusion barrier on the high-temperature steam oxidation of FeCrAl coatings on Zry-4","authors":"Haiyan Liao ,&nbsp;Xiaohan Deng ,&nbsp;Weijiu Huang ,&nbsp;Haibo Ruan ,&nbsp;Shuai Lyu ,&nbsp;Yuan Niu ,&nbsp;Xiangkong Xu ,&nbsp;Yongyao Su ,&nbsp;Junjun Wang","doi":"10.1016/j.jnucmat.2025.156422","DOIUrl":"10.1016/j.jnucmat.2025.156422","url":null,"abstract":"<div><div>This study systematically investigates the effects of Al content (3, 5, and 7 wt%) and a Cr-N diffusion barrier on the high-temperature steam oxidation performance of magnetron-sputtered FeCrAl coatings deposited on Zry-4. The tests were conducted at 1200 °C under simulated loss of coolant accident (LOCA) conditions in a steam environment. Results reveal that a minimum Al content of 5 wt% is required to form a continuous and dense Al₂O₃ layer, which significantly improves oxidation resistance. The Cr-N interlayer effectively suppressed the interdiffusion of Zr and Fe by promoting the in-situ formation of a continuous Zr(Fe,Cr)₂ Laves phase barrier during oxidation. This barrier inhibited the development of low-melting-point Fe-Zr eutectic phases, thereby enhancing interfacial stability. Compared with a pure Cr interlayer, the Cr-N barrier demonstrated superior performance in mitigating interdiffusion and improving coating durability. These findings highlight the critical roles of Al content and the Cr-N interlayer in optimizing FeCrAl coatings for accident-tolerant fuel (ATF) claddings, offering essential insights for designing zirconium alloy coatings with superior oxidation resistance and reliability under extreme conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156422"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882269","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":"Spatially heterogeneous evolution of helium bubbles in He-irradiated Inconel 617: Experimental observation and anisotropic phase-field simulation","authors":"Chong Liu ,&nbsp;Dazhao Cheng ,&nbsp;Jiahui Qu ,&nbsp;Dehui Li ,&nbsp;Yan Zhao ,&nbsp;Jing Zhang","doi":"10.1016/j.jnucmat.2025.156417","DOIUrl":"10.1016/j.jnucmat.2025.156417","url":null,"abstract":"<div><div>The evolution of helium bubbles in nickel-based alloys poses significant challenges to the structural integrity of nuclear reactor components. This study investigates the spatial heterogeneity and anisotropic evolution of helium bubbles in Inconel 617 under He ion irradiation (1 MeV, 3 × 10¹⁶ ions/cm², 830 °C) using a coupled experimental and anisotropic phase-field approach. Transmission electron microscopy reveals distinct bubble characteristics: the (Ni,Cr)O oxide/matrix interface exhibits the highest bubble density, while both the interface and the oxide interior share similar, fine bubble sizes (∼2.4 nm). In contrast, the Ni matrix hosts larger bubbles (∼3.8 nm) at a lower density. The anisotropic phase-field model, developed to quantitatively interpret these observations, identifies the coherent oxide interface as a potent defect sink. Its strong absorption of helium and vacancies promotes dense bubble nucleation while inducing a local vacancy depletion that suppresses coarsening in the oxide. Conversely, in the matrix, long-range diffusion enables the growth of sparse, large bubbles. Critically, the simulation reproduces the experimentally observed quadrilateral bubble morphology in the matrix, confirming the dominance of surface energy anisotropy in shaping bubbles. This work provides the first quantitative interpretation of anisotropic bubble evolution at oxide interfaces in IN617, highlighting the critical interplay between defect sink strength and crystallographic anisotropy. The coupled methodology establishes a predictive framework for designing radiation-tolerant microstructures.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156417"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882273","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":"Inducing δ-ferrite in 316NG austenitic stainless steel without welding: Microstructure and SCC resistance in simulated BWR environment","authors":"Zhao-Wei Fu,&nbsp;Hiroshi Abe,&nbsp;Yutaka Watanabe","doi":"10.1016/j.jnucmat.2025.156383","DOIUrl":"10.1016/j.jnucmat.2025.156383","url":null,"abstract":"<div><div>This study aims to develop a stainless-steel microstructure with enhanced resistance to stress corrosion cracking propagation in high-temperature water by uniformly distributing island shaped δ-ferrite along grain boundaries. A partially melted zone near the fusion boundary was reproduced through weld-simulation heat treatments to investigate the formation behavior of δ-ferrite. The key material factors and optimal heat treatment conditions required to obtain a microstructure with uniformly and widely distributed island-shaped δ-ferrite along grain boundaries were examined. Crack growth tests were then conducted in a simulated boiling water reactor environment to evaluate the effect of δ-ferrite on crack propagation. The effect of the δ-ferrite on the crack growth rate was not clear. This was attributed to the low crack growth rate and the insufficient distribution of δ-ferrite along the crack propagation path. An investigation of all secondary cracks observed in the cross sections that encountered δ-ferrite revealed that, although the volume fraction of δ-ferrite along the total crack length was extremely small, &gt;50% of the crack tips were located at δ-ferrite. This result suggests that δ-ferrite possesses resistance to crack propagation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156383"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838098","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}