Journal of Nuclear Materials最新文献

{"title":"Study on the effect of heterogeneous Pu distribution on the thermal conductivity of MOX fuel","authors":"Koji Kitano ,&nbsp;Yuji Ohishi ,&nbsp;Hiroaki Muta","doi":"10.1016/j.jnucmat.2025.156212","DOIUrl":"10.1016/j.jnucmat.2025.156212","url":null,"abstract":"<div><div>The MOX fuel with heterogeneous plutonium (Pu) distribution is utilized in light water reactors. In some studies, irradiated heterogeneous MOX fuel shows higher thermal conductivity than the irradiated homogeneous MOX fuel. However, the effect of heterogeneous Pu distribution on the thermal conductivity has not been completely understood. In this study, thermal resistivity as well as thermal conductivity is discussed for unirradiated and irradiated MOX fuel with heterogeneous Pu distribution. First, existing model formulas were reviewed to represent the thermal resistivity of homogeneous MOX fuel, considering phonon scattering by Pu atoms, fission products, and irradiation defects. Subsequently, a model formula was proposed to describe the thermal conductivity of heterogeneous MOX fuel based on the concept of thermal resistance circuits that could be considered similar to electrical resistance circuits. The predictions from the proposed model formulas were compared against experimental measurements of thermal conductivity for both homogeneous and heterogeneous MOX fuels. The comparison demonstrated that the model formulas properly predicted the thermal conductivity of both types of MOX fuel. Furthermore, the model formulas suggest that the uranium-rich phase, which contains minimal Pu in the heterogeneous MOX fuel, serves as a fast pathway for phonon conduction. This contributes to the higher thermal conductivity observed in irradiated heterogeneous MOX fuel compared to irradiated homogeneous MOX fuel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156212"},"PeriodicalIF":3.2,"publicationDate":"2025-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321512","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":"Evaluation of tungsten as diffusion barrier for high-density U-Mo fuel using heavy-ion irradiation","authors":"C. Schwarz ,&nbsp;J. Shi ,&nbsp;A. Oberbauer ,&nbsp;B. Stepnik ,&nbsp;W. Petry","doi":"10.1016/j.jnucmat.2025.156210","DOIUrl":"10.1016/j.jnucmat.2025.156210","url":null,"abstract":"<div><div>Monolithic plate-type nuclear fuels for research reactors based on a U-Mo alloy require a coating barrier between the fuel zone and the Al-based cladding to suppress the growth of the undesired interaction layer. The effectiveness of tungsten - an alternative to the commonly used zirconium - is studied to this end. Physical vapor deposition (PVD) serves as a suitable method to produce the required barrier layers. These W coatings turn out brittle, especially when combined with a subsequent cladding application, but show promising behavior in limiting thermal diffusion. This is confirmed in heavy-ion irradiation experiments to simulate in-pile performance of the fuel system. Irradiation of the W barrier from the direction of the Al cladding as well as from the U-Mo side displays mainly interaction of W and Al. As observed, a layer of 1 µm W is sufficient to prevent interaction of U-Mo and Al up to fission density equivalents of 7.5·10²⁰ fissions/cm³. The temperature effect dominates the growth of interaction layer when the sample temperature is increased from 140 °C to 200 °C.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156210"},"PeriodicalIF":3.2,"publicationDate":"2025-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263454","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":"Tritium release performance of centrifugally granulated Li2TiO3 pebbles","authors":"Guangfan Tan ,&nbsp;Yasuhisa Oya ,&nbsp;Yingchun Zhang ,&nbsp;Chang-An Wang ,&nbsp;Yanhao Dong","doi":"10.1016/j.jnucmat.2025.156213","DOIUrl":"10.1016/j.jnucmat.2025.156213","url":null,"abstract":"<div><div>Li₂TiO₃ ceramic pebbles with high chemical stability and excellent mechanical strength are regarded as promising tritium breeding materials for their ensured safety and sustained operation in deuterium-tritium (D-T) fusion reactors. Compared with traditional preparation process, centrifugal granulation method is an ideal technology for preparing tritium breeder due to its advantages such as easy large-scale production, low cost, high sphericity and excellent mechanical properties. However, the tritium release behaviors of Li₂TiO₃ ceramic pebbles prepared by centrifugal granulation are not clear. In this paper, the tritium release behaviors of Li₂TiO₃ ceramic pebbles have been systematically investigated by using thermal desorption spectroscopy (TDS). This will guide the way to adjust the microstructure of Li₂TiO₃ ceramic pebbles in the future. The TDS results indicate that a single desorption peak can be observed at 791K, with HTO being the main release form. The isothermal heating experiments confirmed that the release of tritium in the main form of HTO was governed by diffusion, while the overall tritium release was also influenced by defect trapping and de-trapping mechanisms. Additionally, considering the actual service conditions of the ceramic breeder, the tritium release behaviors of Li₂TiO₃ ceramic pebbles exposed to 900 °C for 50 h and the purge gas with helium were systematically studied, respectively. The experiment results revealed that the release temperature of tritium increased to 827 K, and the helium purge gas had a slight effect on tritium release temperature. Therefore, these results provide guidance to enhance the tritium release amount and rate of Li₂TiO₃ ceramic pebbles in the future.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156213"},"PeriodicalIF":3.2,"publicationDate":"2025-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321509","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":"Investigation of the incorporation capacity of calcine of simulated high-level liquid waste in borosilicate glass matrix","authors":"Wang Changfu ,&nbsp;Liu Xu ,&nbsp;Li Lin ,&nbsp;Li Wengzheng ,&nbsp;Shao Bobo ,&nbsp;Liu Yan ,&nbsp;Wang Yun ,&nbsp;Huang Xi ,&nbsp;Tan Shengheng ,&nbsp;Liu Zhirong ,&nbsp;Zhang Shengdong","doi":"10.1016/j.jnucmat.2025.156208","DOIUrl":"10.1016/j.jnucmat.2025.156208","url":null,"abstract":"<div><div>As a pivotal technology for high-level liquid waste (HLW) treatment, vitrification plays a crucial role in ensuring the sustainable development of nuclear energy. The two-step vitrification process, which involves initial calcination of HLW into solid calcine followed by melting with base glass, has demonstrated significant improvements in immobilization efficiency. This study investigated the capacity of simulated HLW calcine to be incorporated within a borosilicate glass matrix.Through comprehensive characterization techniques including field emission scanning electron microscopy coupled with energy dispersive spectroscopy (SEM-EDS), X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), and Raman spectroscopy, the effects of calcine content (18–32 wt%) on the microstructure, chemical composition, phase distribution, and network structure of the vitrified matrixs were examined. Homogeneous glass matrices are achieved at 18–26 wt% calcine content, while phase separation and crystalline precipitation occur at higher concentrations. However, when the content reaches 28 wt%, the initial precipitation of granular CaMoO₄ crystals occurs, maintaining relatively uniform distribution. Further increasing the content to 30 wt% leads to the formation of acicular CaMoO₄ phases accompanied by minor Zr_xCe_1-xO₂ particle precipitation. Notably, at 32 wt% calcined product content, a distinct phase separation layer emerges on the surface of the vitrified matrix, with a substantial increase in Zr_xCe_1-xO₂ precipitates. These research findings have clarified the incorporation capacity limits of simulated HLW calcine in borosilicate glass matrix, providing valuable theoretical guidance for enhancing waste loading in vitrified forms during the vitrification process.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156208"},"PeriodicalIF":3.2,"publicationDate":"2025-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321510","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":"Automated hydride segmentation and quantification in zirconium alloys using a convolutional neural network-assisted machine learning model","authors":"M.M. Darwhekar ,&nbsp;Y. Pushpalatha Devi ,&nbsp;T. Narayana Murty ,&nbsp;K.V. Mani Krishna ,&nbsp;R.N. Singh","doi":"10.1016/j.jnucmat.2025.156209","DOIUrl":"10.1016/j.jnucmat.2025.156209","url":null,"abstract":"<div><div>Accurate quantification of zirconium hydrides is critical for assessing the structural integrity of zirconium alloys in nuclear reactors. This study presents a machine learning (ML)-based framework for automated hydride segmentation in optical micrographs using a convolutional neural network trained on curated, augmented datasets. The performance is benchmarked against an advanced conventional algorithm employing Contrast-Limited Adaptive Histogram Equalization (CLAHE) -based adaptive thresholding. Results demonstrate superior robustness, consistency, and throughput of the ML approach across varying image qualities. However, the adaptive thresholding method retains advantages in high-fidelity conditions, offering rapid and reproducible segmentation. This work advances scalable hydride quantification, supporting both high-throughput screening and precision lab characterization. The presented workflow is modular and adaptable to a wide range of microstructural segmentation problems, including phase and inclusion analysis.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156209"},"PeriodicalIF":3.2,"publicationDate":"2025-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263738","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":"Irradiation tolerant medium-entropy multilayer films with interface optimization mechanism","authors":"Li Jiang ,&nbsp;Jiannan Hao ,&nbsp;Guanzhi Wang ,&nbsp;Guang Ran ,&nbsp;Yanhui Li ,&nbsp;Qing Peng ,&nbsp;Wei Zhang ,&nbsp;Lumin Wang","doi":"10.1016/j.jnucmat.2025.156207","DOIUrl":"10.1016/j.jnucmat.2025.156207","url":null,"abstract":"<div><div>Irradiation induced helium damage is harmful in nuclear energy, thus irradiation resistant materials are highly desirable. Nanostructuring is an advanced strategy to mitigate the helium damage with pre-exist defects. In this work, we have investigated the irradiation effects in Cu/(Ta₅₀Nb₂₅Mo₂₅) medium-entropy multilayer films using experiments accompanied with molecular dynamics simulations. The multilayers with modulation period of 5 and 50 nm have been irradiated with 60 keV He ions at 300 and 400 °C, respectively, with total fluence of 10²¹ ions/m². The multilayers exhibit commendable microstructural stability, with the four constituent elements maintaining their intended modulation distribution after irradiation. A diverse range of bubble sizes is observed within the Cu layers, with sizes peaking at 3 to 5 nm in the region of maximal damage. However, bubbles with a diameter around 1 nm are detected in the Ta₅₀Nb₂₅Mo₂₅ layers, indicating that the bubbles are in the nucleation phase. The 50 nm multilayer shows two types of layer interfaces, i.e., (01<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>)_BCC || (1<span><math><mrow><mover><mn>1</mn><mo>¯</mo></mover><mn>1</mn></mrow></math></span>)_FCC, [<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>11]_BCC || [011]_FCC (Type 1, K-S relationship), and (01<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>)_BCC || (<span><math><mrow><mover><mn>1</mn><mo>¯</mo></mover><mn>10</mn></mrow></math></span>)_FCC, [<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>11]_BCC || [001]_FCC (Type 2). Bubbles tend to distribute along the Type 1 interface, molecular dynamics results suggest that the average formation energy of vacancy clusters in Type 1 is lower than that in Type 2, which is consistent with the trend of the monovacancy. As a result, the swelling rate of the medium-entropy multilayers is significantly decelerated compared with the bimetal multilayers.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156207"},"PeriodicalIF":3.2,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263453","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":"Reaction between molten 316 L stainless steel and Zircaloy–4 at 1450 ℃ during a core melt accident of LWR","authors":"Tatsuya Kanno ,&nbsp;Takayuki Iwama ,&nbsp;Elizaveta Cheremisina ,&nbsp;Hiroshi Fukaya ,&nbsp;Lichun Zheng ,&nbsp;Ryo Inoue ,&nbsp;Shigeru Ueda","doi":"10.1016/j.jnucmat.2025.156205","DOIUrl":"10.1016/j.jnucmat.2025.156205","url":null,"abstract":"<div><div>Regarding the interaction between Zircaloy and stainless steel in the accidental conditions of light water reactors, the detailed mechanism of the eutectic liquefaction has been widely studied. However, there are seldom data on the interaction between the surface-oxidized Zircaloy and molten stainless steel, which could happen in several potential accident scenarios. The dissolution reaction rate of Zr from the surface-oxidized Zircaloy into the molten SS316L was experimentally evaluated at 1450 °C. The amount of Zr transferred externally by the Zircaloy–SS316L reaction can be formulated as</div><div><span><math><mrow><mfrac><mrow><mi>m</mi><mo>(</mo><mi>x</mi><mo>,</mo><mi>t</mi><mo>)</mo></mrow><mi>A</mi></mfrac><mo>=</mo><mrow><mo>(</mo><mo>−</mo><mn>0.0002</mn><mi>x</mi><mo>+</mo><mn>0.115</mn><mo>)</mo></mrow><mo>·</mo><mrow><mo>{</mo><mtext>ln</mtext><mrow><mo>(</mo><mi>t</mi><mo>+</mo><mn>0.759</mn><mo>)</mo></mrow><mspace></mspace><mo>+</mo><mspace></mspace><mn>0.275</mn><mo>}</mo></mrow><mspace></mspace><mrow><mo>[</mo><mrow><mi>g</mi></mrow><mo>/</mo><msup><mrow><mtext>cm</mtext></mrow><mn>2</mn></msup><mo>]</mo></mrow><mo>,</mo></mrow></math></span></div><div>where <em>m</em>(<em>x,t</em>), <em>A, x</em>, and <em>t</em> represent the amount of transferred Zr [g], reaction area [cm²], thickness of oxide film on Zircaloy surface [µm], and reaction time [min], respectively.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156205"},"PeriodicalIF":3.2,"publicationDate":"2025-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263739","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":"Synthesis of metallic uranium microspheres via electrolytic reduction in molten LiCl","authors":"Han-Hung Hsu ,&nbsp;Gamze Colak ,&nbsp;Gregory Leinders ,&nbsp;Thomas Cardinaels ,&nbsp;Tom Breugelmans ,&nbsp;Bart Geboes","doi":"10.1016/j.jnucmat.2025.156188","DOIUrl":"10.1016/j.jnucmat.2025.156188","url":null,"abstract":"<div><div>This study presents the synthesis of metallic uranium microspheres via a high temperature electrolytic reduction process. A two-step procedure is applied where first UO₂ microspheres are prepared via an internal gelation method and are then applied as feed material for electrolytic reduction in molten LiCl-Li₂O at 923 K. The use of microspheres combines several advantages over traditional powder metallurgical processes requiring powder or pellet feeds. A relatively high active surface area provides sufficient reaction rates while the lack of powder streams enables safer handling and reduces the risk of material loss during electrolysis. X-ray diffraction (XRD) and thermogravimetric analysis (TGA) demonstrated that UO₂ microspheres achieve higher conversion to the metallic form relative to conventional pellets, confirming their superior reduction efficiency. Moreover, the reduced microspheres and pellets were characterized by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDX) to investigate the structural and morphological changes during the reduction process. The reduced uranium microspheres maintained a porous, sponge-like structure, offering better structural integrity and stability, which is advantageous for further processing. Metallic uranium with minor inclusion of impurities was achieved as a final product as confirmed by elemental analysis through inductively coupled plasma mass spectrometry (ICP-MS) indicating impurity levels below 50 ppm of the main impurity elements. Increased concentrations of Mo and Ni were detected originating from electrode contamination after prolonged electrolysis. Li and Al from the residual salt were also detected in the final product. Although further optimization is necessary to fully comply with ASTM standards for uranium metal, this study outlines a promising approach for preparing high-purity metallic uranium from uranyl nitrate solutions with limited process steps and no additional purification, proving potentially useful for applications such as medical isotope production and recycling of target production scraps.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156188"},"PeriodicalIF":3.2,"publicationDate":"2025-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263737","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":"Natural isotopic compositions of titanium, iron, and nickel observed in commercial fuel pellets – Promising candidate elements for stable isotope tagging","authors":"Jan Render,&nbsp;Quinn R. Shollenberger,&nbsp;Naomi Marks,&nbsp;John M. Rolison","doi":"10.1016/j.jnucmat.2025.156200","DOIUrl":"10.1016/j.jnucmat.2025.156200","url":null,"abstract":"<div><div>Stable isotope taggants would constitute unique identifiers for nuclear fuel cycle materials, resulting in expedited timelines and high confidence provenance assessments for nuclear forensics investigations. However, reliably identifying and interpreting stable isotope taggants in nuclear materials recovered from outside of regulatory control will largely be predicated on the assumption that the taggant element intrinsic to the untagged nuclear material exhibits natural isotopic ratios. Here, we present high-precision Ti, Fe, and Ni isotope compositions in 13 commercial low-enriched uranium (LEU) fuel pellets to assess the suitability of these transition metals as stable isotope taggants. Our investigations reveal limited isotope variations among the fuel pellets in all three elements, which are consistent with small mass-dependent isotope fractionations, comparable to variations previously reported for natural samples.</div><div>In practice, isotopically tagged nuclear materials are expected to fall along isotopic mixing lines, since intrinsic background levels of taggant elements dilute the taggant towards natural isotope compositions. The observation that Ti, Fe, and Ni isotope compositions in a suite of LEU fuel pellets are close to or indistinguishable from estimates for the Bulk Silicate Earth demonstrates that a two end-member mixing assumption would be valid for these transition metals, indicating that all three are promising candidate elements for stable isotope tagging. Finally, we present mass balance calculations to quantify isotopic perturbations expected from admixing isotopically anomalous Ti, Fe, and Ni taggants to assess the interplay between elemental and taggant concentrations and find favorable compromises for facilitating successful taggant identification with current analytical methods.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156200"},"PeriodicalIF":3.2,"publicationDate":"2025-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263741","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":"Grain refinement and formation mechanism of cast uranium by multiple β quenching and α annealing treatments","authors":"D.L. Zou ,&nbsp;F. Liu ,&nbsp;J.Q. Yin,&nbsp;S.S. Cui,&nbsp;T. Shi,&nbsp;M.S. Zhai,&nbsp;D.L. Chen,&nbsp;A.Y. Yin,&nbsp;K.Z. Liu,&nbsp;W. Shi","doi":"10.1016/j.jnucmat.2025.156203","DOIUrl":"10.1016/j.jnucmat.2025.156203","url":null,"abstract":"<div><div>The grain refinement and formation mechanism of cast uranium by multiple β quenching and α annealing treatments were systematically investigated. The results show that the grain refinement of cast uranium was maximised by multiple β quenching and α annealing treatments, with the final grain size refined to approximately 100 μm and becoming stable after two thermal cycles. After β quenching, high density low angle grain boundaries were formed in grains, and decreased significantly after α annealing due to the recrystallization. The formation mechanism of the grain refinement of cast uranium was mainly attributed to the deformation recrystallization. Large lattice distortion caused by β to α phase transformation, as well as contraction anisotropy, was assisted with high temperature annealing and led to the occurrence of the recrystallization to form the fine equiaxed crystallized grains. The surface natural oxidation rates and colors of the fine-grained uranium were closely related with multiple factors, mainly including grain microhardness, grain orientation, oxidation inhomogeneity and defect density.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156203"},"PeriodicalIF":3.2,"publicationDate":"2025-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263458","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}