Journal of Nuclear Materials最新文献_第4页

Investigations of size effect and normalization models for tensile deformation of A508-III steel miniature specimens A508-III钢微缩试样拉伸变形尺寸效应及归一化模型研究

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-08 DOI: 10.1016/j.jnucmat.2026.156446

Cong Chen , Qi Zhang , Shengmin Xin , Wanhuan Yang , Junwan Li , Weihua Zhong , Guangsheng Ning

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 λ 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.

通过改变A508-III钢的热处理温度，制备了不同晶粒尺寸和厚度的微型试样。对微缩试样进行了室温拉伸试验，分析了拉伸尺寸效应的机理。结果表明，厚度和晶粒尺寸的变化对拉伸性能有显著的尺寸效应。通过引入综合表征微观试样晶粒效应和特征尺寸效应的影响参数λ，建立了考虑尺寸效应的Swift力学本构模型。为了完善微型试件的损伤模型，采用有限元辅助试验（FAT）方法确定了微型试件的延性损伤演化参数。基于建立的力学本构模型和延性损伤模型，通过有限元分析得到的微型试件载荷-位移曲线与试验试验结果的平均误差小于5%。采用有限元法预测了微型试样的拉伸力学性能，建立了屈服和抗拉强度的归一化模型。

{"title":"Investigations of size effect and normalization models for tensile deformation of A508-III steel miniature specimens","authors":"Cong Chen , Qi Zhang , Shengmin Xin , Wanhuan Yang , Junwan Li , Weihua Zhong , 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":"2026-01-08","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}

引用次数: 0

Homogenization-based irradiation performance analysis of U-Mo/Al dispersion fuel plates 基于均匀化的U-Mo/Al弥散燃料板辐照性能分析

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-07 DOI: 10.1016/j.jnucmat.2026.156442

Wei Li , Haokun Wang , Jie Wang , Qi Lu , Yuanming Li , Junmei Wu , Wenjie Zeng

The U-Mo/Al dispersion fuel has emerged as a highly promising candidate for high-density Low Enriched Uranium (LEU) fuel in high-performance research reactors. Understanding the irradiation behavior of U-Mo/Al dispersion fuel plates is crucial for ensuring the safety and performance of these reactors. In this study, we propose a homogenization-based approach to analyze the irradiation performance of U-Mo/Al dispersion fuel plates. At the mesoscopic scale, the Representative Volume Element (RVE) approach, combined with the Finite Element Method (FEM), is employed to determine the equivalent thermomechanical properties of the dispersion fuel meat. To account for the significant influence of the U-Mo/Al interaction layer (IL), the empirical model of the growth of this layer under irradiation is seamlessly integrated into the RVE model. Subsequently, based on the obtained equivalent thermomechanical property models, irradiation performance analysis of U-Mo/Al fuel plate at the macroscopic scale is conducted. Both the meso‑ and macroscale models have been validated against publicly available experimental data. By informing macroscopic performance analysis with mesoscopic homogenization, this study provides a reliable and efficient tool for optimizing the design and enhancing the performance of advanced dispersion fuels in research reactor applications.

U-Mo/Al弥散燃料已成为高性能研究堆高密度低浓铀（LEU）燃料的极具前景的候选材料。了解铀钼/铝弥散燃料板的辐照行为对确保这些反应堆的安全和性能至关重要。在这项研究中，我们提出了一种基于均质化的方法来分析U-Mo/Al分散燃料板的辐照性能。在细观尺度上，采用代表性体积元法（RVE）和有限元法（FEM）相结合的方法确定了分散燃料肉的等效热力学性能。考虑到U-Mo/Al相互作用层（IL）的显著影响，将该层在辐照下生长的经验模型无缝集成到RVE模型中。随后，基于得到的等效热力学模型，对U-Mo/Al燃料板进行了宏观尺度的辐照性能分析。中尺度和宏观尺度模型都已根据公开的实验数据进行了验证。通过介观均质为宏观性能分析提供信息，本研究为研究堆应用中先进分散燃料的优化设计和提高性能提供了可靠而有效的工具。

{"title":"Homogenization-based irradiation performance analysis of U-Mo/Al dispersion fuel plates","authors":"Wei Li , Haokun Wang , Jie Wang , Qi Lu , Yuanming Li , Junmei Wu , Wenjie Zeng","doi":"10.1016/j.jnucmat.2026.156442","DOIUrl":"10.1016/j.jnucmat.2026.156442","url":null,"abstract":"<div><div>The U-Mo/Al dispersion fuel has emerged as a highly promising candidate for high-density Low Enriched Uranium (LEU) fuel in high-performance research reactors. Understanding the irradiation behavior of U-Mo/Al dispersion fuel plates is crucial for ensuring the safety and performance of these reactors. In this study, we propose a homogenization-based approach to analyze the irradiation performance of U-Mo/Al dispersion fuel plates. At the mesoscopic scale, the Representative Volume Element (RVE) approach, combined with the Finite Element Method (FEM), is employed to determine the equivalent thermomechanical properties of the dispersion fuel meat. To account for the significant influence of the U-Mo/Al interaction layer (IL), the empirical model of the growth of this layer under irradiation is seamlessly integrated into the RVE model. Subsequently, based on the obtained equivalent thermomechanical property models, irradiation performance analysis of U-Mo/Al fuel plate at the macroscopic scale is conducted. Both the meso‑ and macroscale models have been validated against publicly available experimental data. By informing macroscopic performance analysis with mesoscopic homogenization, this study provides a reliable and efficient tool for optimizing the design and enhancing the performance of advanced dispersion fuels in research reactor applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156442"},"PeriodicalIF":3.2,"publicationDate":"2026-01-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024748","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

Corrigendum to “Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography” [Journal of Nuclear Materials Volume 620 (2026) 156341] “使用三维x射线计算机断层扫描AGR-3/4 TRISO燃料致密体辐照后检查”的勘误表[核材料杂志第620卷（2026）156341]

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-07 DOI: 10.1016/j.jnucmat.2025.156396

Swapnil Morankar , William C. Chuirazzi , Rahul R. Kancharla , Josh J. Kane , Brian Gross , Nikolaus L. Cordes , John D. Stempien

引用次数: 0

Creep properties of solution-annealed 316H stainless steel made by laser-powder directed energy deposition additive manufacturing 激光粉末定向能沉积增材制造316H不锈钢固溶退火的蠕变性能

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-07 DOI: 10.1016/j.jnucmat.2026.156444

John Snitzer , Benjamin Sutton , John Shingledecker , Xiaoyuan Lou

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.

这项工作研究了激光粉末定向能沉积（LP-DED）生产的溶液退火增材制造（AM） 316H不锈钢（SS）的高温蠕变响应，并支持将AM用于先进核应用的努力。研究了激光功率和构筑方向对材料蠕变性能的影响。在650℃下，在130 ~ 200 MPa的不同应力下进行蠕变试验，以确保位错爬升为主的蠕变是主要机制。固溶退火去除了原有的化学偏析和胞状结构，形成了非均质、部分再结晶的晶粒结构和由氧化物颗粒固定的亚晶粒结构。尽管溶液退火，激光功率和构建方向显著影响AM试样的微观结构，导致AM试样表现出更高的最小蠕变速率和更低的断裂寿命，与本工作中测试的变形316H相比；然而，AM样品仍然在现有的锻造316H数据库的散射带内。构建方向对蠕变延展性有显著影响，水平构建的试件与垂直构建和变形的试件相比，蠕变延展性降低，这是由于晶界主要垂直于加载方向。蠕变空化主要沿晶界观察到Cr23C6和η-氮化物析出。在AM试样中也观察到晶内析出，但由于极低的面积密度，对蠕变行为的影响有限。幂律分析证实位错爬升是蠕变的主要机制，AM的指数为6.1 ~ 6.9，锻造的指数为8.4。使用Larson-Miller和Monkman-Grant方法进行的蠕变寿命分析显示，加工变异性会导致蠕变寿命的变化；然而，am生产的材料再次落在或高于变形散射带。总体而言，结果表明，固溶退火的DED AM 316H SS可以获得与变形合金一致的蠕变性能，支持其在高温结构应用中的潜在考虑。

{"title":"Creep properties of solution-annealed 316H stainless steel made by laser-powder directed energy deposition additive manufacturing","authors":"John Snitzer , Benjamin Sutton , John Shingledecker , 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 Cr<sub>23</sub>C<sub>6</sub> 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":"2026-01-07","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}

引用次数: 0

A deep learning interatomic potential model for the generation and evolution mechanisms of irradiation defects in yttrium hydride 氢化钇辐照缺陷产生与演化机制的深度学习原子间势模型

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-06 DOI: 10.1016/j.jnucmat.2026.156439

Xi Wang , Meng Tang , Biao Xu , Ji-Jung Kai , Wangyu Hu , Huiqiu Deng

Yttrium hydride (YH₂) serves as a critical high-temperature moderator in advanced reactors, yet its performance under irradiation remains poorly understood. We have developed a high-accuracy deep potential for YH₂ with short-range corrections and validated its reliability in describing defect properties. Using this potential, we carried out threshold displacement energy calculations, which represent a critical irradiation parameter, together with cascade collision simulations. Our results reveal a pronounced elemental disparity in threshold displacement energy, with hydrogen exhibiting a mean value of 5.21 eV compared to 20.86 eV for yttrium, indicating significantly greater susceptibility to displacement. Defect formation probability calculations confirm that hydrogen dominates in initial damage production. Although large scale cascade simulations show complex damage evolution with substantial defect recombination during thermal spike quenching, persistent hydrogen vacancies constitute the majority of residual damage. The higher recovery rate of Y vacancies (∼80%) versus H vacancies (∼70%) reflects different recombination kinetics arising from the contrasting mobilities and lattice environments of Y (on fcc lattice sites) and H (on interstitial sites). These findings provide atomic scale insights into radiation damage behavior in YH₂, highlighting the critical role of H vacancy accumulation in long-term degradation of nuclear moderator materials.

氢化钇（YH2）在先进反应堆中是一种关键的高温慢化剂，但其在辐照下的性能尚不清楚。我们开发了具有短程校正的YH2的高精度深电位，并验证了其在描述缺陷特性方面的可靠性。利用这个势，我们进行了阈值位移能量计算，它代表了一个关键的辐照参数，并进行了级联碰撞模拟。我们的研究结果揭示了在阈值位移能上的明显元素差异，氢的平均值为5.21 eV，而钇的平均值为20.86 eV，这表明氢对位移的敏感性明显更高。缺陷形成概率计算证实氢在初始损伤产生中占主导地位。大尺度叶栅模拟显示，在热刺淬火过程中，损伤演化过程复杂，存在大量缺陷复合，但残余损伤主要是持续的氢空位。Y空位的回收率（~ 80%）高于H空位的回收率（~ 70%），这反映了Y（在fcc点位上）和H（在间隙点位上）不同的迁移率和晶格环境所引起的不同的重组动力学。这些发现为YH2的辐射损伤行为提供了原子尺度上的见解，突出了H空位积累在核慢化材料长期降解中的关键作用。

{"title":"A deep learning interatomic potential model for the generation and evolution mechanisms of irradiation defects in yttrium hydride","authors":"Xi Wang , Meng Tang , Biao Xu , Ji-Jung Kai , Wangyu Hu , Huiqiu Deng","doi":"10.1016/j.jnucmat.2026.156439","DOIUrl":"10.1016/j.jnucmat.2026.156439","url":null,"abstract":"<div><div>Yttrium hydride (YH<sub>2</sub>) serves as a critical high-temperature moderator in advanced reactors, yet its performance under irradiation remains poorly understood. We have developed a high-accuracy deep potential for YH<sub>2</sub> with short-range corrections and validated its reliability in describing defect properties. Using this potential, we carried out threshold displacement energy calculations, which represent a critical irradiation parameter, together with cascade collision simulations. Our results reveal a pronounced elemental disparity in threshold displacement energy, with hydrogen exhibiting a mean value of 5.21 eV compared to 20.86 eV for yttrium, indicating significantly greater susceptibility to displacement. Defect formation probability calculations confirm that hydrogen dominates in initial damage production. Although large scale cascade simulations show complex damage evolution with substantial defect recombination during thermal spike quenching, persistent hydrogen vacancies constitute the majority of residual damage. The higher recovery rate of Y vacancies (∼80%) versus H vacancies (∼70%) reflects different recombination kinetics arising from the contrasting mobilities and lattice environments of Y (on fcc lattice sites) and H (on interstitial sites). These findings provide atomic scale insights into radiation damage behavior in YH<sub>2</sub>, highlighting the critical role of H vacancy accumulation in long-term degradation of nuclear moderator materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156439"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024746","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

How the presence of molten fluoride salt influences the tellurization of UNS N10003 alloy? 熔盐的存在如何影响UNS N10003合金的碲化？

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-06 DOI: 10.1016/j.jnucmat.2026.156438

Rui Wu , Anbo Ge , Xin Du , Yafei Wang , Shaoqiang Guo

Tellurium-induced corrosion in molten salt reactors is strongly influenced by the chemistry of tellurium in molten fluoride salts. In our previous work, Ellingham diagram and E-pTe²⁻ diagrams were constructed to illustrate the theoretical differences between the metal tellurization in tellurium vapor and molten fluoride environments. Here we further validate the electrochemistry of the nickel tellurization reaction in molten fluoride media and highlight the variation of alloy tellurization between the metal tellurization in tellurium vapor and molten fluoride environments. Unlike tellurization in vapor atmosphere, the tellurization of UNS N10003 alloy in molten LiF-NaF-KF eutectic salt shows a strong dependence on electrode potential and only occurs above the Ni/Ni_0.6Te_0.4 reaction potential (1.43 ± 0.03 V). The presence of CrF₃ in the salt influences the tellurium chemistry, thereby accelerating the tellurization of UNS N10003 alloy compared to that observed in the vapor environment.

熔盐堆中碲致腐蚀受熔氟盐中碲的化学性质的强烈影响。在我们之前的工作中，构建了Ellingham图和E-pTe2 -图来说明碲蒸气和熔融氟环境中金属碲化的理论差异。本研究进一步验证了熔氟介质中镍碲化反应的电化学性质，并强调了合金碲化在碲蒸气和熔氟环境中金属碲化的差异。与蒸汽气氛下的碲化不同，UNS N10003合金在熔融的LiF-NaF-KF共晶盐中的碲化表现出强烈的电极电位依赖性，只发生在Ni/Ni0.6Te0.4反应电位（1.43±0.03 V）以上。与蒸汽环境相比，盐中CrF3的存在影响了碲的化学性质，从而加速了UNS N10003合金的碲化。

引用次数: 0

Sustainable granite encapsulation of volatile nuclear wastes: A flux-mediated low-temperature strategy 挥发性核废料的可持续花岗岩封装：一种通量介导的低温策略

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-06 DOI: 10.1016/j.jnucmat.2026.156441

Wenhong Han , Xiaoyan Shu , Wencai Cheng , Guilin Wei , Yuxuan He , Ran Tan , Mingfen Wen , Xirui Lu

To minimize the loss of complex low-melting-point volatile nuclear waste (primarily containing Cs, Sr, Ba, etc.) during the solidification process, B₂O₃ was employed as an auxiliary agent to achieve low-temperature immobilization in amphibolite granulite. The phase evolution, microstructure, elemental distribution, and glass-to-crystal transition of the resulting matrices were systematically characterized. Furthermore, the underlying mechanism by which structural changes of element B in the solidified form affects its properties was explored through density functional theory (DFT) calculations. The results demonstrate that the incorporation of B₂O₃ effectively reduces the sintering temperature by 300°C (from 1500°C to 1200°C). The obtained solidified form exhibits excellent durability, with normalized leaching rates of Cs, Sr, and Ba reaching as low as ∼10⁻³ g·m⁻²·d⁻¹ after 42 days. The formation of a CsAlSiO₄ crystalline phase contributes to retarding the leaching kinetics of radionuclides, thus enhancing the chemical durability. In summary, utilizing B₂O₃ as an auxiliary agent enables amphibolite granulite to leverage the synergistic advantages of glass and ceramic phases, effectively immobilizing volatile TRPO-CS waste (with retention rates of approximately 95% for Cs and 99% for Sr). This strategy provides a promising pathway for the safe and green disposal of radioactive waste.

为了减少复杂低熔点挥发性核废料（主要含有Cs、Sr、Ba等）在固化过程中的损失，采用B2O3作为助剂在角闪石麻粒岩中实现低温固定化。系统地表征了基体的相演化、微观结构、元素分布和玻璃-晶体转变。此外，通过密度泛函理论（DFT）计算，探讨了B元素凝固态结构变化影响其性能的潜在机制。结果表明，B2O3的加入可有效降低烧结温度300℃（从1500℃降至1200℃）。得到的凝固形式表现出优异的耐久性，在42天后，Cs， Sr和Ba的标准化浸出率低至~ 10⁻3 g·m⁻2·d⁻1。CsAlSiO4晶相的形成有助于延缓放射性核素的浸出动力学，从而提高化学耐久性。综上所述，使用B2O3作为助剂，可以使角闪岩麻粒岩利用玻璃相和陶瓷相的协同优势，有效地固定挥发性TRPO-CS废物（Cs的保留率约为95%，Sr的保留率约为99%）。该策略为安全、绿色处置放射性废物提供了一条有希望的途径。

{"title":"Sustainable granite encapsulation of volatile nuclear wastes: A flux-mediated low-temperature strategy","authors":"Wenhong Han , Xiaoyan Shu , Wencai Cheng , Guilin Wei , Yuxuan He , Ran Tan , Mingfen Wen , Xirui Lu","doi":"10.1016/j.jnucmat.2026.156441","DOIUrl":"10.1016/j.jnucmat.2026.156441","url":null,"abstract":"<div><div>To minimize the loss of complex low-melting-point volatile nuclear waste (primarily containing Cs, Sr, Ba, etc.) during the solidification process, B<sub>2</sub>O<sub>3</sub> was employed as an auxiliary agent to achieve low-temperature immobilization in amphibolite granulite. The phase evolution, microstructure, elemental distribution, and glass-to-crystal transition of the resulting matrices were systematically characterized. Furthermore, the underlying mechanism by which structural changes of element B in the solidified form affects its properties was explored through density functional theory (DFT) calculations. The results demonstrate that the incorporation of B<sub>2</sub>O<sub>3</sub> effectively reduces the sintering temperature by 300°C (from 1500°C to 1200°C). The obtained solidified form exhibits excellent durability, with normalized leaching rates of Cs, Sr, and Ba reaching as low as ∼10⁻<sup>3</sup> g·m⁻<sup>2</sup>·d⁻<sup>1</sup> after 42 days. The formation of a CsAlSiO<sub>4</sub> crystalline phase contributes to retarding the leaching kinetics of radionuclides, thus enhancing the chemical durability. In summary, utilizing B<sub>2</sub>O<sub>3</sub> as an auxiliary agent enables amphibolite granulite to leverage the synergistic advantages of glass and ceramic phases, effectively immobilizing volatile TRPO-CS waste (with retention rates of approximately 95% for Cs and 99% for Sr). This strategy provides a promising pathway for the safe and green disposal of radioactive waste.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156441"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922721","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

Fabrication of UO2–Mo composite fuel pellets with enhanced thermal conductivity by using wet mixing 湿法混合制备增强导热性的UO2-Mo复合燃料球团

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-05 DOI: 10.1016/j.jnucmat.2026.156440

Fihan Alharbi, Hywel Ragnauth, Timothy Abram, Joel Turner

Uranium dioxide (UO₂) is the standard nuclear fuel for light water reactors (LWRs) due to its operational experience, irradiation stability, and ease of manufacture. However, its low thermal conductivity results in a high temperature gradient across the pellet in-service, leading to thermal stresses, deformation, and cracking. This study explores the addition of a high thermal conductivity molybdenum phase as a nano-powder, which is trialled alongside pre-sintered UO₂ granules to improve the thermal conductivity of the resulting pellet. UO₂- 10 wt. % Mo composite pellets were fabricated by dispersing UO₂ granules and nano-Mo powder in ethanol during mixing, followed by ethanol evaporation and subsequent Spark Plasma Sintering (SPS) at 1473 K with a 5-minute hold. Pellet microstructures were characterized using scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pellet thermal conductivity was measured by the laser flash method. Significant improvements in thermal conductivity were observed in the as-manufactured pellets with increases of up to 75% at 1073K for the pellets produced from nano-Mo and pre-sintered UO₂ granules compared to a pure UO₂ pellet. These results highlight the effectiveness of nano-Mo addition and the pre-sintering of UO₂ granules in enhancing the thermal performance of UO₂-based nuclear fuel composites.

二氧化铀（UO₂）是轻水反应堆（LWRs）的标准核燃料，因为它的运行经验，辐射稳定性和易于制造。然而，它的低导热性导致在使用中的球团温度梯度高，导致热应力、变形和开裂。本研究探索了高导热钼相作为纳米粉末的添加，并与预烧结的UO2颗粒一起试验，以提高所得颗粒的导热性。UO2- 10 wt. % Mo复合球团制备UO2- 10 wt. % Mo复合球团的方法是将UO2颗粒和纳米Mo粉末分散在乙醇中，然后进行乙醇蒸发，随后在1473 K下进行放电等离子烧结（SPS），保温5分钟。采用扫描电子显微镜（SEM）和x射线衍射仪（XRD）对颗粒的微观结构进行了表征。采用激光闪光法测量颗粒导热系数。与纯UO2颗粒相比，纳米mo和预烧结UO2颗粒生产的颗粒在1073K下的导热性显著提高，增加了75%。这些结果突出了纳米mo添加和UO2颗粒预烧结在提高UO₂基核燃料复合材料热性能方面的有效性。

{"title":"Fabrication of UO2–Mo composite fuel pellets with enhanced thermal conductivity by using wet mixing","authors":"Fihan Alharbi, Hywel Ragnauth, Timothy Abram, Joel Turner","doi":"10.1016/j.jnucmat.2026.156440","DOIUrl":"10.1016/j.jnucmat.2026.156440","url":null,"abstract":"<div><div>Uranium dioxide (UO₂) is the standard nuclear fuel for light water reactors (LWRs) due to its operational experience, irradiation stability, and ease of manufacture. However, its low thermal conductivity results in a high temperature gradient across the pellet in-service, leading to thermal stresses, deformation, and cracking. This study explores the addition of a high thermal conductivity molybdenum phase as a nano-powder, which is trialled alongside pre-sintered UO<sub>2</sub> granules to improve the thermal conductivity of the resulting pellet. UO<sub>2</sub>- 10 wt. % Mo composite pellets were fabricated by dispersing UO₂ granules and nano-Mo powder in ethanol during mixing, followed by ethanol evaporation and subsequent Spark Plasma Sintering (SPS) at 1473 K with a 5-minute hold. Pellet microstructures were characterized using scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pellet thermal conductivity was measured by the laser flash method. Significant improvements in thermal conductivity were observed in the as-manufactured pellets with increases of up to 75% at 1073K for the pellets produced from nano-Mo and pre-sintered UO<sub>2</sub> granules compared to a pure UO₂ pellet. These results highlight the effectiveness of nano-Mo addition and the pre-sintering of UO<sub>2</sub> granules in enhancing the thermal performance of UO₂-based nuclear fuel composites.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156440"},"PeriodicalIF":3.2,"publicationDate":"2026-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976100","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

Origin of the inner-layer stratification of 316 L in 430℃ high-temperature steam 430℃高温蒸汽中316l内层分层的成因

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-03 DOI: 10.1016/j.jnucmat.2026.156437

Haodong Wu , Yaqing Ren , Xiangguo Li , Jian Xu

The oxide stratification behavior of the oxide on 316 L stainless steel (SS) in high-temperature steam at 430 °C with different dissolved oxygen (DO) concentrations (<5/50/200 ppb) was systematically studied. The oxide layer has a double-layer structure. However, the stratification of the inner layer is related to DO. In the high DO environment (50/200 ppb), the Cr-Fe oxide layer exhibited a certain element stratification, owing to the combined action of interstitial hydrogen (H_i) and O. DFT calculations showed that the diffusion barrier of H_i in FeOOH was higher than that in Fe₂O₃ and Cr₂O₃. In a low-DO environment (<5 ppb), the transition from internal to external oxidation is driven by the lack of oxygen. The Cr content in the Cr-rich layer was significantly higher than that in the matrix. The distribution of Ni in the Ni-enriched layer was related to that of the Cr-rich layer.

系统研究了316l不锈钢（SS）在430℃高温蒸汽中不同溶解氧（DO）浓度（<5/50/200 ppb）下氧化物的氧化分层行为。氧化层具有双层结构。然而，内层的分层与DO有关。在高DO环境下（50/200 ppb），由于间隙氢（Hi）和氧的共同作用，Cr-Fe氧化物层呈现出一定的元素分层。DFT计算表明，Hi在FeOOH中的扩散势垒高于Fe2O3和Cr2O3。在低do环境（<5 ppb）中，从内部氧化到外部氧化的转变是由缺氧驱动的。富Cr层中Cr含量显著高于基体中Cr含量。富镍层中Ni的分布与富cr层的分布有关。

引用次数: 0

Effect of Mo/Nb addition on recrystallization behavior and high temperature mechanical properties of FeCrAl alloy tubes Mo/Nb添加对FeCrAl合金管的再结晶行为和高温力学性能的影响

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2026-01-02 DOI: 10.1016/j.jnucmat.2026.156435

Zhe Liu , Zhihao Wang , Ding Zuo , Wenbo Liu , Huiqun Liu , Ruiqian Zhang

The microstructure and high temperature mechanical properties of annealed FeCrAl-Mo/Nb alloy were studied in this paper. The effects of alloying elements Mo and Nb on the microstructure, recrystallization behavior and mechanical properties of FeCrAl alloy were systematically analyzed, and the mechanism was discussed. The results show that the recrystallization behavior of FeCrAl alloy with high Mo content is significantly delayed because more solid solution atoms hinder the dislocation movement. The recrystallization behavior of FeCrAl-2Mo0.65Nb alloy is promoted by the particles stimulated recrystallization nucleation due to the existence of Laves phase at the initial stage of recrystallization. At the later stage of recrystallization, the recrystallized grains of this alloy are not easy to grow due to the pinning effect of Laves phase on the grain boundary, and the average grain size of fully recrystallized grains is small, which is 6.51 μm. The main recrystallization mechanism of FeCrAl alloy is strain induced grain boundary migration nucleation and growth. The addition of Mo and Nb has no obvious effect on the recrystallization texture type of FeCrAl alloy, but it would form different maximum texture strength. And the room temperature hardness of the alloy is improved by solution strengthening and second phase strengthening, respectively. The contribution of Laves phase to the high temperature strength of FeCrAl alloy is limited. The FeCrAl-2Mo0.65Nb alloy with partially recrystallized microstructure shows relatively good strength and ductility at 600 °C. Due to the existence of high-density dislocations and Laves phase, the alloy has relatively large displacement and deceleration creep time at 400 °C.

研究了退火后的feral - mo /Nb合金的显微组织和高温力学性能。系统分析了合金元素Mo和Nb对FeCrAl合金组织、再结晶行为和力学性能的影响，并探讨了其作用机理。结果表明，高Mo含量的FeCrAl合金的再结晶行为明显延迟，因为更多的固溶体原子阻碍了位错的移动。FeCrAl-2Mo0.65Nb合金的再结晶行为是由再结晶初期Laves相的存在引起的颗粒激发的再结晶成核促进的。在再结晶后期，由于Laves相在晶界上的钉钉作用，合金的再结晶晶粒不易长大，完全再结晶晶粒的平均晶粒尺寸较小，为6.51 μm。FeCrAl合金的再结晶机制主要是应变诱导晶界迁移、形核和长大。Mo和Nb的加入对FeCrAl合金的再结晶织构类型没有明显影响，但会形成不同的最大织构强度。通过固溶强化和第二相强化分别提高了合金的室温硬度。Laves相对FeCrAl合金高温强度的贡献是有限的。部分再结晶组织的feral - 2mo0.65 nb合金在600℃时表现出较好的强度和塑性。由于高密度位错和Laves相的存在，合金在400℃时具有较大的位移和减速蠕变时间。

{"title":"Effect of Mo/Nb addition on recrystallization behavior and high temperature mechanical properties of FeCrAl alloy tubes","authors":"Zhe Liu , Zhihao Wang , Ding Zuo , Wenbo Liu , Huiqun Liu , Ruiqian Zhang","doi":"10.1016/j.jnucmat.2026.156435","DOIUrl":"10.1016/j.jnucmat.2026.156435","url":null,"abstract":"<div><div>The microstructure and high temperature mechanical properties of annealed FeCrAl-Mo/Nb alloy were studied in this paper. The effects of alloying elements Mo and Nb on the microstructure, recrystallization behavior and mechanical properties of FeCrAl alloy were systematically analyzed, and the mechanism was discussed. The results show that the recrystallization behavior of FeCrAl alloy with high Mo content is significantly delayed because more solid solution atoms hinder the dislocation movement. The recrystallization behavior of FeCrAl-2Mo0.65Nb alloy is promoted by the particles stimulated recrystallization nucleation due to the existence of Laves phase at the initial stage of recrystallization. At the later stage of recrystallization, the recrystallized grains of this alloy are not easy to grow due to the pinning effect of Laves phase on the grain boundary, and the average grain size of fully recrystallized grains is small, which is 6.51 μm. The main recrystallization mechanism of FeCrAl alloy is strain induced grain boundary migration nucleation and growth. The addition of Mo and Nb has no obvious effect on the recrystallization texture type of FeCrAl alloy, but it would form different maximum texture strength. And the room temperature hardness of the alloy is improved by solution strengthening and second phase strengthening, respectively. The contribution of Laves phase to the high temperature strength of FeCrAl alloy is limited. The FeCrAl-2Mo0.65Nb alloy with partially recrystallized microstructure shows relatively good strength and ductility at 600 °C. Due to the existence of high-density dislocations and Laves phase, the alloy has relatively large displacement and deceleration creep time at 400 °C.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156435"},"PeriodicalIF":3.2,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922718","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