Journal of Nuclear Materials最新文献

{"title":"Influence of multi-ion concentrations on corrosion product deposition on zirconium alloy cladding in high-temperature water","authors":"Yu Yang ,&nbsp;Meijiao Huang ,&nbsp;Jixue Sui ,&nbsp;An Li ,&nbsp;Guangming Shen ,&nbsp;Xiaoyong Wu ,&nbsp;Lu Wu ,&nbsp;Mingzhang Lin","doi":"10.1016/j.jnucmat.2025.156420","DOIUrl":"10.1016/j.jnucmat.2025.156420","url":null,"abstract":"<div><div>The concentrations of soluble metal ions in the primary water circulation loop of pressurized water reactors can greatly affect the deposition behaviors of corrosion products on zirconium alloy cladding tubes at high temperatures, which is referred to as CRUD. However, while considerable progress has been made in the development of CRUD deposition models, the applicability of these models remains limited, and the impacts of Fe³⁺, Ni²⁺, Cr³⁺, Co²⁺, and Mn²⁺ on the microstructure of CRUD and the corrosion mechanism have not been fully incorporated into model development. The present work addresses these issues by combining experimental characterization with thermodynamic calculations to analyze the mechanism by which high-temperature aqueous solutions with different Fe³⁺, Ni²⁺, Cr³⁺, Co²⁺, and Mn²⁺ ion concentrations influence the CRUD deposition process on zirconium alloy cladding tubes. In addition to the composition and structure of CRUD deposits, the oxide film properties of the cladding tubes are further examined under subcooled nucleate boiling (SNB) and non-SNB conditions. The results indicate that increases in the weight of zirconium alloy cladding tubes under the non-SNB condition are due primarily to oxide film thickening under low metal ion concentrations. At high ion concentrations, a thinner oxide film was observed, which is likely attributed to modifications in the local chemical environment at the metal/oxide interface induced by the CRUD deposits. In contrast, the CRUD morphology changed under the SNB condition from cluster-like deposits without chimneys to porous structures with chimneys with increasing metal ion concentrations. Experimental results obtained under the SNB condition, in addition to the results of thermodynamic calculations of the Gibbs free energy, demonstrate that Fe₃O₄ and spinel compounds, such as CoFe₂O₄ and FeCr₂O₄, are the dominant phases within the CRUD deposits.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156420"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882262","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":"Pyrophoricity of uranium and uranium compounds: Mechanisms, knowledge gaps, and implications for nuclear safety","authors":"Alisha J. Cramer ,&nbsp;Peter G. Martin ,&nbsp;Thomas B. Scott","doi":"10.1016/j.jnucmat.2025.156425","DOIUrl":"10.1016/j.jnucmat.2025.156425","url":null,"abstract":"<div><div>Depleted uranium materials pose a significant challenge with respect to their safe, long-term storage, not only from a radiological standpoint, but also with regards to fire safety owing to the reactivity of finely-divided uranium material with air. The risk of pyrophoric ignition is known to be strongly dependent on the material’s specific surface area, however, other factors are known to likely play a role in either increasing or decreasing such a risk. In this article, the research to date on the pyrophoric behaviour of uranium compounds is compiled and examined with a view to understanding the factors controlling pyrophoric ignition and determining the gaps in current knowledge. Although some influencing factors, such as specific surface area, dominate ignition behaviour, several other factors can disrupt expected behaviour, demonstrating that pyrophoric ignition is a complex, dynamic interplay of different factors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156425"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882267","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":"Protective Al-rich oxide scale formation in low-Cr alumina-forming martensitic steels under liquid lead corrosion","authors":"Cesar Fernandez-Jimenez ,&nbsp;Isaac Toda-Caraballo ,&nbsp;Roger Castellote-Alvarez ,&nbsp;Peter Szakalos ,&nbsp;Christopher Petersson ,&nbsp;José A. Jiménez ,&nbsp;Carlos Capdevila ,&nbsp;David San-Martin","doi":"10.1016/j.jnucmat.2026.156468","DOIUrl":"10.1016/j.jnucmat.2026.156468","url":null,"abstract":"<div><div>Two alumina-forming martensitic (AFM) steels containing less than 8 wt.% Cr were exposed for over 1000 h at 550 °C and 650 °C in static liquid Pb with controlled oxygen concentration to evaluate the formation of an alumina oxide scale and its protective capacity against corrosion. The AFM-1 steel (3.6 wt.% Al, 7.8 wt.% Cr) formed a continuous, protective oxide scale that effectively resisted Pb penetration under all conditions, particularly at 650 °C, where performance improved due to the dissolution of B2-NiAl precipitates during prolonged exposure, releasing Al that migrated to the surface and enabled the formation of an Al-rich oxide layer, ensuring sustained protection and mitigating the localized nodular oxidation observed at 550 °C. In contrast, the AFM-2 steel (2.9 wt.% Al, 7.5 wt.% Cr) failed to develop a complete protective oxide layer, allowing molten Pb to penetrate and react with the substrate even at 650 °C, causing severe oxidation. These results demonstrate that, beyond a high Ni content (12 wt.%), achieving excellent corrosion resistance in liquid Pb requires a synergistic combination of an optimal Al/Cr balance, dissolution of B2-NiAl precipitates as an Al source, and enhanced atomic diffusion facilitated by the high density of subgrain boundaries in the martensitic microstructure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156468"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024743","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 electrochemical corrosion potential in PWR coolant for advanced FeCrAl and Zr–4 alloys via radiolysis and mixed potential modelling","authors":"Zijian Lin,&nbsp;Changjiang Hu,&nbsp;Yifan Li,&nbsp;Sinan Chen,&nbsp;Minzhang Lin,&nbsp;Jun Ma","doi":"10.1016/j.jnucmat.2026.156451","DOIUrl":"10.1016/j.jnucmat.2026.156451","url":null,"abstract":"<div><div>The predication of advanced alloy corrosion risk by the H₂O₂ generation during coolant radiolysis has been an important issue in Pressurized water reactor (PWR) maintenance but remains rarely explored. Herein, this study developed the radiolysis and mixed-potential modelling (MPM) based on ⁶⁰Co γ-irradiation products analysis together with electrochemical corrosion experiments conducted under non-irradiated conditions. First, the experimental data correlates the H₂O₂ production with a wide array of influencing practical factors such as temperature, LiOH/NH₃/H₃BO₃ concentrations, providing a radiolysis model for corrosion basis. We next conducted the electrochemical corrosion behavior of FeCrAl (APMT and C26M) and Zircaloy–4 (Zr–4) alloys in simulated PWR coolants to develop MPM. The modeling predicts that hydrogen injections can reduce the Electrochemical Corrosion Potential (ECP). Under identical condition, Zr–4 exhibits the most positive ECP, while APMT shows the most negative value. Threshold analysis suggests that injecting approximately 5 mL (STP)/kg H₂ can lower the ECP of all tested alloys below −0.23 V (vs. SHE) that can prevent stress corrosion cracking risk.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156451"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976101","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 morphological response of plasma-facing tungsten: Effects of surface crystallographic orientation and prediction of fuzz onset","authors":"Omeet N. Patel ,&nbsp;Dwaipayan Dasgupta ,&nbsp;Asanka Weerasinghe ,&nbsp;Brian D. Wirth ,&nbsp;Dimitrios Maroudas","doi":"10.1016/j.jnucmat.2026.156450","DOIUrl":"10.1016/j.jnucmat.2026.156450","url":null,"abstract":"<div><div>Recent experiments have revealed a strong dependence of surface morphology on surface crystallographic orientation during the initial phase of ‘fuzz’ growth in fusion plasma-facing tungsten. Here, we examine this dependence using predictions of dynamical simulations based on an atomistically-informed continuum-scale model of surface evolution in plasma-irradiated tungsten (W). Upon exposure to a low-energy helium plasma, tungsten surface adatoms are produced as a result of surface vacancy-adatom pair formation and the flux of tungsten self-interstitial atoms, resulting from growing over-pressurized helium bubbles, toward the plasma-exposed surface. A combination of a stress-governed surface morphological instability, the preferential diffusion of tungsten surface adatoms, and the surface free energy anisotropy leads to the growth of various features on the plasma-facing surface. Simulations of the morphological response of helium irradiated W(111), W(110), and W(100) surfaces show the growth of triangular, stripe-shaped, and pyramidal surface features, respectively, and predict the growth kinetics, namely, the thickness evolution of a damaged surface layer at the early stage of fuzz formation, consistent with experimental observations. This work elucidates the effect of surface crystallographic orientation on the morphological evolution of plasma-facing tungsten surfaces, where the stress-governed surface morphological instability drives the formation of surface features (stripes or mounds), the preferential surface adatom migration controls the anisotropic growth of these features, while the surface free energy anisotropy is responsible for the formation of planar facets on the features emanating from the plasma-facing surface. We observe that the average separation between the resulting surface features increases with increasing temperature and duration of plasma exposure. This coarsening process is thermally activated, with the temperature dependence of the average feature separation well described by an Arrhenius relationship. Our modeling framework enables a predictive assessment of fuzz onset in plasma-facing tungsten across different crystallographic orientations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156450"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976102","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 and mechanism of Fe³⁺ impurities on the corrosion behavior of SiC-B₄C composites in molten FLiNaK salt","authors":"Congcong Zhao ,&nbsp;Junhong Jia ,&nbsp;Kui Wang ,&nbsp;Xiaoyu Jiao ,&nbsp;Runze Wei ,&nbsp;Jian Wang ,&nbsp;Jie Yang ,&nbsp;Yaoqiang Tang ,&nbsp;Qingqing Zhang ,&nbsp;Zongyu Zhang","doi":"10.1016/j.jnucmat.2026.156447","DOIUrl":"10.1016/j.jnucmat.2026.156447","url":null,"abstract":"<div><div>This study systematically investigates the critical influence of Fe³⁺ impurity ions on the corrosion behavior of boron carbide (B₄C) -reinforced silicon carbide (SiC) composite ceramics in FLiNaK molten salt. SiC-B₄C composites were fabricated via hot-press sintering, and their corrosion characteristics were compared in both original and Fe³⁺-contaminated environments. The results indicate that in pure molten salt, the addition of B₄C promotes the formation of a dense, graphitized carbon layer on the surface, which significantly enhances corrosion resistance. In contrast, the strong oxidizing power of Fe³⁺ disrupts the formation and stability of this protective layer. This disruption results in a loose, porous surface structure, a substantial increase in corrosion layer thickness and the creation of pathways for the corrosive melt penetration. Thermodynamic analysis further confirms that FeF₃, as a potent oxidizing agent, preferentially reacts with B₄C. Moreover, Fe³⁺ is regenerated through a cyclic disproportionation reaction (3Fe²⁺→2Fe³⁺+Fe), which intensifies the corrosion process. This work provides critical theoretical support for material selection and service-life assessment of structural components in molten salt reactors under practical operating conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156447"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957740","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":"In – situ chlorination of simulated fission products from light water reactor fuel using ZrCl4","authors":"Diallo Barnes,&nbsp;Courtney Eckley,&nbsp;Mary Cernyar,&nbsp;Michael F. Simpson","doi":"10.1016/j.jnucmat.2025.156418","DOIUrl":"10.1016/j.jnucmat.2025.156418","url":null,"abstract":"<div><div>To develop a process for chlorinating spent nuclear fuel (SNF) from light water reactors and dissolving it into a molten salt, the reactivity of several surrogate fission products (SFPs) with NiCl₂ and ZrCl₄ in molten LiCl – KCl was tested. ZrCl₄ was created <em>in-situ</em> by reacting NiCl₂ and Zr metal in the molten salt. Powders of SrO, La₂O₃, CeO₂, Cs₂O, and a Mo metal rod were all immersed in eutectic LiCl – KCl with initially 8.9 wt% NiCl₂, likely exceeding the solubility limit. Salt samples were collected at several time intervals over 24 h and analyzed via inductively coupled plasma mass spectrometry (ICP – MS). NiCl₂ did not directly react with any of the SFPs, but <em>in situ</em> formed ZrCl₄ was reactive towards all the oxides. The Mo metal rod was virtually inert towards reaction with both NiCl₂ and ZrCl₄. Experiments were run at 773 or 873 K with or without stirring condition. Conversion of SrO, CeO₂, and La₂O₃ to soluble chlorides ranged from 87 – 93 %, while Cs₂O conversion was only 64 % on average. Some of the Ni metal by-product plated onto the Mo rod, which was unreactive towards chlorination. A solid containing Ni and ZrO₂ settled to the bottom of the crucible.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156418"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922725","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":"Molecular dynamics study on the effect of interstitial hydrogen clusters on the slip of edge dislocation in tungsten","authors":"Yifan Liu ,&nbsp;Ke Xu ,&nbsp;Zhiye Tang ,&nbsp;Chenxin Yan ,&nbsp;Shuo Jin ,&nbsp;Linyun Liang ,&nbsp;Hong-Bo Zhou ,&nbsp;Guang-Hong Lu","doi":"10.1016/j.jnucmat.2026.156448","DOIUrl":"10.1016/j.jnucmat.2026.156448","url":null,"abstract":"<div><div>Interstitial hydrogen (H) clusters with rock salt structure exhibit energy stability in tungsten (W) and play a crucial role in enhancing its hardness. However, the underlying physical mechanisms and the specific hardening behavior remain unclear. To this end, we systematically investigate the effect of the H cluster on the slip behavior of an 1/2 [111] (<span><math><mrow><mn>1</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>0</mn></mrow></math></span>) edge dislocation in W by using the molecular dynamics method. We first study the slip of the edge dislocation in the absence of the H cluster, which reveals typical phonon drag control characteristics. Based on this, a slip drag coefficient <em>B</em>(<em>T</em>) is obtained, enabling accurate prediction of the dislocation mobility under various temperatures and stresses. In the presence of H clusters, the hardening effect in W is significantly enhanced. Notably, the geometric parameters of the H cluster, i.e., height and diameter, exert significant regulatory influence on the slip behavior of the edge dislocation through a quantitative correlation. Furthermore, the critical resolved shear stress (CRSS) displays a slight dependence on temperature within the range of 100 K-800 K, indicating that the dislocation motion is primarily governed by the geometry of the H cluster. These results provide new insights into the mechanisms of H-induced irradiation hardening in W, offering valuable data to support the development of high-performance W-based materials with enhanced irradiation resistance and long-term service stability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156448"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976097","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":"Influence of defect sink strength on dislocation loop and helium bubble evolution in low-alloy steel under in-situ He irradiation and annealing","authors":"Jiyong Huang ,&nbsp;Hucheng Yu ,&nbsp;Yifan Ding ,&nbsp;Xiangbing Liu ,&nbsp;Ziqi Cao ,&nbsp;Runzhong Wang ,&nbsp;Wenqing Jia ,&nbsp;Sheng Fan ,&nbsp;Guang Ran","doi":"10.1016/j.jnucmat.2026.156461","DOIUrl":"10.1016/j.jnucmat.2026.156461","url":null,"abstract":"<div><div>Understanding the formation and evolution of irradiation-induced defects under different effective sink strengths is essential for designing radiation-resistant structural materials in advanced nuclear systems. In this study, in situ He ion irradiation experiments were conducted at 450 °C on two regions of low-alloy steel materials with distinct sink strengths characterized by variations in dislocation density and precipitate distribution. In situ TEM observations revealed the formation and evolution behaviors of dislocation loops and helium bubbles during irradiation and subsequent annealing at 700 and 750 °C. Regions with higher sink strength exhibited lower dislocation loop density, smaller loop size, and a reduced fraction of <em>a</em> &lt; 100&gt; loops, indicating enhanced point-defect recombination and suppressed loop nucleation and growth. During irradiation, helium bubbles in the low-sink region were larger and non-spherical, while those in the high-sink region were finer and denser. Upon annealing, bubble size reversal occurred: bubbles in the high-sink region grew rapidly at later stages. At the final annealing stage, bubbles in the high-sink region were largely pinned, whereas smaller bubbles in the low-sink region exhibited thermally driven migration. These findings elucidate the critical role of effective sink strength in defect evolution and provide experimental insights for the microstructural design of radiation-resistant materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156461"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976098","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":"Creep properties of solution-annealed 316H stainless steel made by laser-powder directed energy deposition additive manufacturing","authors":"John Snitzer ,&nbsp;Benjamin Sutton ,&nbsp;John Shingledecker ,&nbsp;Xiaoyuan Lou","doi":"10.1016/j.jnucmat.2026.156444","DOIUrl":"10.1016/j.jnucmat.2026.156444","url":null,"abstract":"<div><div>This work examines the high-temperature creep response of solution-annealed additively manufactured (AM) 316H stainless steel (SS) produced by laser-powder directed energy deposition (LP-DED) and support the efforts towards using AM for advanced nuclear applications. The work focused on the effects of laser power and build direction on creep properties. Creep testing was conducted at 650 °C at varying stresses from 130 to 200 MPa to ensure that dislocation climb-dominated creep was the primary mechanism. Solution annealing removed the as-built chemical segregation and cellular structures, forming a heterogeneous, partially recrystallized grain structure with subgrain structures pinned by oxide particles. Despite solution annealing, the laser power and build direction significantly influenced the microstructure resulting in AM specimens exhibiting higher minimum creep rates and lower rupture lives compared to wrought 316H tested in this work; however, the AM specimens remained within the scatterband of an existing wrought 316H database. The build direction significantly impacted the creep ductility, where the horizontally built specimens exhibited reduced creep ductility compared to vertically built and wrought specimens, attributed to the grain boundaries being primarily oriented perpendicular to the loading direction. Creep cavitation was observed primarily along grain boundaries near Cr₂₃C₆ and η-nitride precipitates. Intragranular precipitates were also observed in AM specimens but, due to the very low area density, had a limited effect on creep behavior. Power law analysis confirmed dislocation climb was the dominant creep mechanism in all specimens with exponents ranging from 6.1 to 6.9 for AM and 8.4 for wrought. Creep life analysis, conducted using Larson-Miller and Monkman-Grant methods, revealed processing variability induced varying creep life; however, the AM-produced material again fell within or higher than the wrought scatterband. Overall, the results indicate that solution-annealed, DED AM 316H SS can achieve creep performance consistent with wrought alloys, supporting its potential consideration for high-temperature structural applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156444"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976103","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}