Pub Date : 2026-01-22DOI: 10.1016/j.jnucmat.2026.156475
R.S. Stroud , C. Reynolds , T. Melichar , J. Haley , M. Carter , M. Moody , C. Hardie , D. Bowden , D. Nguyen-Manh , M.R. Wenman
VN precipitates used to strengthen ARAFM steels for fusion applications dissolve under high Fe ion irradiation (100 dpa at dpa · s, 600 ∘C). This study examined point defects and solute substitutions using atom probe tomography, machine learning interatomic potentials, and density functional theory. Combined with transmission electron microscopy, results show N-vacancies and substitutional Cr exist in VN precipitates before irradiation. Monte Carlo simulations and collision cascade simulations confirm ordered vacancies at operating temperatures help mitigate irradiation damage. However, solute additions disrupt vacancy ordering and enhance irradiation-induced damage, potentially accelerating dissolution.
{"title":"Defects and impurity properties of VN precipitates in ARAFM steels: Modelling using a universal machine learning potential and experimental validation","authors":"R.S. Stroud , C. Reynolds , T. Melichar , J. Haley , M. Carter , M. Moody , C. Hardie , D. Bowden , D. Nguyen-Manh , M.R. Wenman","doi":"10.1016/j.jnucmat.2026.156475","DOIUrl":"10.1016/j.jnucmat.2026.156475","url":null,"abstract":"<div><div>VN precipitates used to strengthen ARAFM steels for fusion applications dissolve under high Fe ion irradiation (100 dpa at <span><math><msup><mn>10</mn><mrow><mo>−</mo><mn>3</mn></mrow></msup></math></span> dpa · s<span><math><msup><mrow></mrow><mrow><mo>−</mo><mn>1</mn></mrow></msup></math></span>, 600 <sup>∘</sup>C). This study examined point defects and solute substitutions using atom probe tomography, machine learning interatomic potentials, and density functional theory. Combined with transmission electron microscopy, results show N-vacancies and substitutional Cr exist in VN precipitates before irradiation. Monte Carlo simulations and collision cascade simulations confirm ordered vacancies at operating temperatures help mitigate irradiation damage. However, solute additions disrupt vacancy ordering and enhance irradiation-induced damage, potentially accelerating dissolution.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"624 ","pages":"Article 156475"},"PeriodicalIF":3.2,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146071107","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}
Pub Date : 2026-01-20DOI: 10.1016/j.jnucmat.2026.156473
R. Juneja , M. Wirtz , J.W. Coenen , J. Rapp , E.A. Unterberg
Fusion materials in magnetic fusion devices must tolerate extreme loading conditions. Tungsten remains the leading plasma-facing component (PFC) material, yet the microstructural variability governing its degradation under harsh conditions, including crack initiation, porosity evolution, and thermal fatigue, remains costly to characterize at scale. We developed a data-efficient generative workflow to synthesize realistic scanning electron microscopy (SEM) microstructures of tungsten for physics-aware data augmentation and rapid hypothesis testing under fusion-relevant environmental conditions. Starting from 3200 SEMs acquired from electron-beam (e-beam) heat flux exposure studies on tungsten, we tiled each image into fixed-field 256 × 256 grayscale patches and trained two models: a baseline model, conditional-GAN (c-GAN) and a style-based model, conditional StyleGAN2 with Adaptive Discriminator Augmentation (c-StyleGAN2-ADA). The latter adapts augmentation strength during training and is well suited to the small-data regime. Fidelity was assessed with a physics-aware validation suite: (i) image distributional similarity to real SEMs via Fréchet Inception Distance (FID) and Kernel Inception Distance (KID); (ii) microstructure realism via grain-size statistics derived from classical image analysis and comparison of grain-area statistics via Kolmogorov-Smirnov (KS) and Earth Mover’s Distance (EMD); and (iii) anti-memorization screening using nearest-neighbor searches with Learned Perceptual Image Patch Similarity (LPIPS). Notably, our trained c-StyleGAN2-ADA generator reproduced grain-size distributions that closely followed the real data while maintaining diversity and avoiding trivial copies, outperforming the c-GAN on both perceptual and physics-aware metrics. The approach yields physically plausible microstructures on demand and provides a basis to seed multi-scale degradation models, uncertainty analyses, and “virtual experiments” for PFC design when direct measurements are scarce.
{"title":"Machine learning for microstructure synthesis in fusion materials: A physics-aware validation framework for tungsten plasma-facing components","authors":"R. Juneja , M. Wirtz , J.W. Coenen , J. Rapp , E.A. Unterberg","doi":"10.1016/j.jnucmat.2026.156473","DOIUrl":"10.1016/j.jnucmat.2026.156473","url":null,"abstract":"<div><div>Fusion materials in magnetic fusion devices must tolerate extreme loading conditions. Tungsten remains the leading plasma-facing component (PFC) material, yet the microstructural variability governing its degradation under harsh conditions, including crack initiation, porosity evolution, and thermal fatigue, remains costly to characterize at scale. We developed a data-efficient generative workflow to synthesize realistic scanning electron microscopy (SEM) microstructures of tungsten for physics-aware data augmentation and rapid hypothesis testing under fusion-relevant environmental conditions. Starting from 3200 SEMs acquired from electron-beam (e-beam) heat flux exposure studies on tungsten, we tiled each image into fixed-field 256 × 256 grayscale patches and trained two models: a baseline model, conditional-GAN (c-GAN) and a style-based model, conditional StyleGAN2 with Adaptive Discriminator Augmentation (c-StyleGAN2-ADA). The latter adapts augmentation strength during training and is well suited to the small-data regime. Fidelity was assessed with a physics-aware validation suite: (i) image distributional similarity to real SEMs via Fréchet Inception Distance (FID) and Kernel Inception Distance (KID); (ii) microstructure realism via grain-size statistics derived from classical image analysis and comparison of grain-area statistics via Kolmogorov-Smirnov (KS) and Earth Mover’s Distance (EMD); and (iii) anti-memorization screening using nearest-neighbor searches with Learned Perceptual Image Patch Similarity (LPIPS). Notably, our trained c-StyleGAN2-ADA generator reproduced grain-size distributions that closely followed the real data while maintaining diversity and avoiding trivial copies, outperforming the c-GAN on both perceptual and physics-aware metrics. The approach yields physically plausible microstructures on demand and provides a basis to seed multi-scale degradation models, uncertainty analyses, and “virtual experiments” for PFC design when direct measurements are scarce.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156473"},"PeriodicalIF":3.2,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146075089","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}
Metal tritides have long been applied in tritium storage due to their high capacity and stability. The decay of tritium produces helium-3 (³He), which is mainly retained in metal tritides in the form of bubbles. Although the evolution of helium-3 bubbles in metal tritides has been of wide concern for a long time, the trend of their morphological transformation is still under debate. In this work, the shape evolution of helium bubbles in typical metal tritides (erbium, titanium, and zirconium) was tracked by transmission electron microscopy. The results show that in the tritides of erbium and titanium, helium-3 bubbles undergo a sphere-to-platelet transformation at the early stage (³He/M = 0.02∼0.06), while in zirconium tritide the helium-3 bubbles remain spherical up to ³He/M > 0.29. Compared with theoretical models, it is found that large and plate-like bubbles can maintain stability by widening rather than through a spherical transformation. Our results further suggest that the dominant energy contribution of helium-3 bubbles shifts from surface energy to strain energy with aging. Overall, the present work provides strong experimental support for investigating helium bubble behavior within metal tritide lattices, offering guidance for the rational design of tritium storage and fusion materials.
{"title":"Spontaneous shape transformation of helium bubble in metal tritide lattice: sphere to platelet","authors":"Muhong Li, Lin Qi, Chengqin Zou, Shuanglin Hu, Weidu Wang, Xiaochun Han, Xiaosong Zhou, Shuming Peng, Huahai Shen","doi":"10.1016/j.jnucmat.2026.156470","DOIUrl":"10.1016/j.jnucmat.2026.156470","url":null,"abstract":"<div><div>Metal tritides have long been applied in tritium storage due to their high capacity and stability. The decay of tritium produces helium-3 (³He), which is mainly retained in metal tritides in the form of bubbles. Although the evolution of helium-3 bubbles in metal tritides has been of wide concern for a long time, the trend of their morphological transformation is still under debate. In this work, the shape evolution of helium bubbles in typical metal tritides (erbium, titanium, and zirconium) was tracked by transmission electron microscopy. The results show that in the tritides of erbium and titanium, helium-3 bubbles undergo a sphere-to-platelet transformation at the early stage (³He/M = 0.02∼0.06), while in zirconium tritide the helium-3 bubbles remain spherical up to ³He/M > 0.29. Compared with theoretical models, it is found that large and plate-like bubbles can maintain stability by widening rather than through a spherical transformation. Our results further suggest that the dominant energy contribution of helium-3 bubbles shifts from surface energy to strain energy with aging. Overall, the present work provides strong experimental support for investigating helium bubble behavior within metal tritide lattices, offering guidance for the rational design of tritium storage and fusion materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156470"},"PeriodicalIF":3.2,"publicationDate":"2026-01-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024751","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}
Pub Date : 2026-01-18DOI: 10.1016/j.jnucmat.2026.156472
Te Zhu , Yuanhang Xia , Qianqian Wang , Ping Fan , Qiaoli Zhang , Hailiang Ma , Daqing Yuan , Baoyi Wang , Qiu Xu , Xingzhong Cao
We investigated the irradiation damage characteristics and mechanical properties of a low-activation high-entropy alloy (LAHEA) with a body-centered cubic structure, which is reinforced with copper nanoparticles (Cu NPs). The average size of the Cu NPs was 26 nm, and they were uniformly dispersed within the BCC matrix, exhibiting perfect coherence with the matrix. The addition of Cu enhanced the yield strength of the alloy from 767 to 1255 MPa and increased the elongation to 15.4%, which is attributed to the pinning effect of the Cu particles on dislocations and their high deformability. After high-dose Ni-ion irradiation at 580°C, selected to align with BCC alloy swelling behavior, the material demonstrated exceptional radiation resistance characterized by the complete suppression of void formation. Cross-sectional microstructural analysis revealed that irradiation-induced Ti-rich semi-coherent precipitates contributed to hardening, while coherent Fe/Mn-rich clusters acted as efficient point-defect recombination centers. These findings indicate that the low-activation HEA has substantial potential for nuclear applications.
{"title":"Enhanced mechanical properties of a copper nanoparticle-reinforced low-activation high-entropy alloy and its implications for irradiation resistance","authors":"Te Zhu , Yuanhang Xia , Qianqian Wang , Ping Fan , Qiaoli Zhang , Hailiang Ma , Daqing Yuan , Baoyi Wang , Qiu Xu , Xingzhong Cao","doi":"10.1016/j.jnucmat.2026.156472","DOIUrl":"10.1016/j.jnucmat.2026.156472","url":null,"abstract":"<div><div>We investigated the irradiation damage characteristics and mechanical properties of a low-activation high-entropy alloy (LAHEA) with a body-centered cubic structure, which is reinforced with copper nanoparticles (Cu NPs). The average size of the Cu NPs was 26 nm, and they were uniformly dispersed within the BCC matrix, exhibiting perfect coherence with the matrix. The addition of Cu enhanced the yield strength of the alloy from 767 to 1255 MPa and increased the elongation to 15.4%, which is attributed to the pinning effect of the Cu particles on dislocations and their high deformability. After high-dose Ni-ion irradiation at 580°C, selected to align with BCC alloy swelling behavior, the material demonstrated exceptional radiation resistance characterized by the complete suppression of void formation. Cross-sectional microstructural analysis revealed that irradiation-induced Ti-rich semi-coherent precipitates contributed to hardening, while coherent Fe/Mn-rich clusters acted as efficient point-defect recombination centers. These findings indicate that the low-activation HEA has substantial potential for nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156472"},"PeriodicalIF":3.2,"publicationDate":"2026-01-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024742","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}
Pub Date : 2026-01-18DOI: 10.1016/j.jnucmat.2026.156471
Guangfan Tan , Xudong Mao , Yasuhisa Oya , Yingchun Zhang , Chang-An Wang , Yanhao Dong
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.
{"title":"Low-temperature fabrication of Li4SiO4 ceramic pebbles with excellent tritium release performance using SiC as silicon source","authors":"Guangfan Tan , Xudong Mao , Yasuhisa Oya , Yingchun Zhang , Chang-An Wang , Yanhao Dong","doi":"10.1016/j.jnucmat.2026.156471","DOIUrl":"10.1016/j.jnucmat.2026.156471","url":null,"abstract":"<div><div>Lithium orthosilicate (Li<sub>4</sub>SiO<sub>4</sub>) 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 Li<sub>4</sub>SiO<sub>4</sub> 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 Li<sub>4</sub>SiO<sub>4</sub> ceramic pebbles at low temperature. The experimental results reveal that the phase formation temperature of the Li<sub>4</sub>SiO<sub>4</sub> could be reduced to 500 °C. When sintered at 850 °C, Li<sub>4</sub>SiO<sub>4</sub> 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 Li₄SiO<sub>4</sub> prepared via the centrifugal granulation method, the samples exhibited rapid tritium desorption rate of 1.1 × 10<sup>5</sup> Bq⋅g<sup>−1</sup>⋅s<sup>−1</sup> 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 Li<sub>4</sub>SiO<sub>4</sub> 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":"2026-01-18","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}
Pub Date : 2026-01-17DOI: 10.1016/j.jnucmat.2026.156466
Xuanning Ma , Zilin Gao , Junhao Xu , Zhongdi Yu , Huan Li , Lianwen Wang , Jianping Xu
The effect of chromium content on corrosion kinetics of nickel-based alloys was investigated using full-immersion tests and potentiodynamic polarization. Results demonstrate that increasing chromium (Cr) content reduces corrosion rates while shifting surface morphology from uniform to localized corrosion. XPS and EBSD analyses reveal that elevated Cr promotes duplex structure formation, generating porous Cr₂O₃ phases at phase boundaries. Fluoride ions (F⁻) penetrate along grain boundaries, inducing intergranular corrosion by preferential dissolution of Cr. Furthermore, Pourbaix diagrams thermodynamically confirm F⁻-accelerated dissolution of Cr/Ni oxides in acidic environments, exacerbating alloy corrosion.
{"title":"Corrosion mechanism of Ni-Cr alloy in boiling fluorine-containing nitric acid: Effect of chromium content","authors":"Xuanning Ma , Zilin Gao , Junhao Xu , Zhongdi Yu , Huan Li , Lianwen Wang , Jianping Xu","doi":"10.1016/j.jnucmat.2026.156466","DOIUrl":"10.1016/j.jnucmat.2026.156466","url":null,"abstract":"<div><div>The effect of chromium content on corrosion kinetics of nickel-based alloys was investigated using full-immersion tests and potentiodynamic polarization. Results demonstrate that increasing chromium (Cr) content reduces corrosion rates while shifting surface morphology from uniform to localized corrosion. XPS and EBSD analyses reveal that elevated Cr promotes duplex structure formation, generating porous Cr₂O₃ phases at phase boundaries. Fluoride ions (F⁻) penetrate along grain boundaries, inducing intergranular corrosion by preferential dissolution of Cr. Furthermore, Pourbaix diagrams thermodynamically confirm F⁻-accelerated dissolution of Cr/Ni oxides in acidic environments, exacerbating alloy corrosion.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156466"},"PeriodicalIF":3.2,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024744","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}
Pub Date : 2026-01-17DOI: 10.1016/j.jnucmat.2026.156467
Elina Charatsidou , Anita Pazzaglia , Kaitlyn Bullock , Maria Giamouridou , Eleanor Lawrence Bright , Mikael Jolkkonen , Christoph Hennig , Pär Olsson
Uranium mononitride (UN) is a promising candidate fuel for next-generation fast reactors due to its high fissile density, superior thermal conductivity, and high melting point compared to conventional oxide fuels. However, scarce experimental data on UN and its thermophysical behaviour under fission product incorporation limits its performance assessment. Zirconium nitride (ZrN) is an efficient thermal conductor and a candidate material for inert matrix fuels. Given its high thermal conductivity, ZrN addition at sufficient concentrations should, in principle, induce percolation conduction and increase thermal conductivity in UN. To decouple chemistry from irradiation-induced porosity, known to dominate thermal degradation at high burnup, this study isolates the intrinsic chemical contribution of Zr incorporation under dense, low-porosity conditions. (U,Zr)N pellets with 6.5 and 20 at. % Zr were fabricated by spark plasma sintering (SPS), using powders produced from arc-melted alloy via the hydride-nitride-denitride route. Synchrotron powder X-ray diffraction confirmed the formation of solid solutions and enhanced Zr solubility after sintering, resulting in improved microstructural homogeneity. Thermal diffusivity was measured between 300 and 1500 K using light flash analysis, and thermal conductivity was derived using heat capacity and density correlations with porosity correction. Despite the intrinsically higher thermal conductivity of ZrN, the incorporation of 6.5 at. % Zr reduced the thermal conductivity relative to UN, consistent with impurity scattering. The 20 at. % Zr composition further decreased conductivity, indicating the microstructure does not meet the conditions required for percolation conduction. Differences in the temperature dependence of thermal diffusivity between UN and Zr-bearing samples highlight a compositional influence on heat transport. The results provide benchmark data for (U,Zr)N and insights into chemical and thermophysical interactions in nitride ceramics.
{"title":"Impact of zirconium incorporation on the thermophysical properties of uranium mononitride","authors":"Elina Charatsidou , Anita Pazzaglia , Kaitlyn Bullock , Maria Giamouridou , Eleanor Lawrence Bright , Mikael Jolkkonen , Christoph Hennig , Pär Olsson","doi":"10.1016/j.jnucmat.2026.156467","DOIUrl":"10.1016/j.jnucmat.2026.156467","url":null,"abstract":"<div><div>Uranium mononitride (UN) is a promising candidate fuel for next-generation fast reactors due to its high fissile density, superior thermal conductivity, and high melting point compared to conventional oxide fuels. However, scarce experimental data on UN and its thermophysical behaviour under fission product incorporation limits its performance assessment. Zirconium nitride (ZrN) is an efficient thermal conductor and a candidate material for inert matrix fuels. Given its high thermal conductivity, ZrN addition at sufficient concentrations should, in principle, induce percolation conduction and increase thermal conductivity in UN. To decouple chemistry from irradiation-induced porosity, known to dominate thermal degradation at high burnup, this study isolates the intrinsic chemical contribution of Zr incorporation under dense, low-porosity conditions. (U,Zr)N pellets with 6.5 and 20 at. % Zr were fabricated by spark plasma sintering (SPS), using powders produced from arc-melted alloy via the hydride-nitride-denitride route. Synchrotron powder X-ray diffraction confirmed the formation of solid solutions and enhanced Zr solubility after sintering, resulting in improved microstructural homogeneity. Thermal diffusivity was measured between 300 and 1500 K using light flash analysis, and thermal conductivity was derived using heat capacity and density correlations with porosity correction. Despite the intrinsically higher thermal conductivity of ZrN, the incorporation of 6.5 at. % Zr reduced the thermal conductivity relative to UN, consistent with impurity scattering. The 20 at. % Zr composition further decreased conductivity, indicating the microstructure does not meet the conditions required for percolation conduction. Differences in the temperature dependence of thermal diffusivity between UN and Zr-bearing samples highlight a compositional influence on heat transport. The results provide benchmark data for (U,Zr)N and insights into chemical and thermophysical interactions in nitride ceramics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156467"},"PeriodicalIF":3.2,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024752","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}
Beryllium metal is characterized by its unique physical properties, which determines its wide range of applications, including the use in nuclear reactors, resulting inevitably in activated metallic beryllium that has to be treated as radioactive waste. In the present work, the corrosion behavior of metallic beryllium in aqueous NaOH solutions with pH ranging between 6.7 and 14.0 and in solutions simulating the environment in potential waste encapsulation matrices such as Ordinary Portland Cement (OPC) or magnesium phosphate cement (MPC) was studied in detail. Corrosion rates of metallic beryllium samples were experimentally studied by using two direct methods based on gravimetric measurements and the determination of beryllium concentrations in the solution by using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). A combined method based on these two direct methods is proposed to enable the determination of corrosion rates in various aqueous solutions, including alkaline solutions and those with near neutral pH values. Detailed studies of corroded metal surfaces were carried out using scanning electron microscopy (SEM) combined with energy dispersive X-ray spectroscopy (EDS), indicating pitting corrosion as prominent corrosion mechanism.
{"title":"Corrosion of metallic beryllium in various aqueous solutions","authors":"Andrey Bukaemskiy , Guido Deissmann , Sebastien Caes , Giuseppe Modolo , Dirk Bosbach","doi":"10.1016/j.jnucmat.2026.156465","DOIUrl":"10.1016/j.jnucmat.2026.156465","url":null,"abstract":"<div><div>Beryllium metal is characterized by its unique physical properties, which determines its wide range of applications, including the use in nuclear reactors, resulting inevitably in activated metallic beryllium that has to be treated as radioactive waste. In the present work, the corrosion behavior of metallic beryllium in aqueous NaOH solutions with pH ranging between 6.7 and 14.0 and in solutions simulating the environment in potential waste encapsulation matrices such as Ordinary Portland Cement (OPC) or magnesium phosphate cement (MPC) was studied in detail. Corrosion rates of metallic beryllium samples were experimentally studied by using two direct methods based on gravimetric measurements and the determination of beryllium concentrations in the solution by using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). A combined method based on these two direct methods is proposed to enable the determination of corrosion rates in various aqueous solutions, including alkaline solutions and those with near neutral pH values. Detailed studies of corroded metal surfaces were carried out using scanning electron microscopy (SEM) combined with energy dispersive X-ray spectroscopy (EDS), indicating pitting corrosion as prominent corrosion mechanism.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156465"},"PeriodicalIF":3.2,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024668","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}
Pub Date : 2026-01-17DOI: 10.1016/j.jnucmat.2026.156468
Cesar Fernandez-Jimenez , Isaac Toda-Caraballo , Roger Castellote-Alvarez , Peter Szakalos , Christopher Petersson , José A. Jiménez , Carlos Capdevila , David San-Martin
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.
{"title":"Protective Al-rich oxide scale formation in low-Cr alumina-forming martensitic steels under liquid lead corrosion","authors":"Cesar Fernandez-Jimenez , Isaac Toda-Caraballo , Roger Castellote-Alvarez , Peter Szakalos , Christopher Petersson , José A. Jiménez , Carlos Capdevila , 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":"2026-01-17","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}
Zr-added FeCrAl oxide-dispersion-strengthened (ODS) alloys have been developed as promising candidates for accident-tolerant fuel claddings, which can spontaneously form protective alumina scales at elevated temperatures and provide excellent oxidation resistance in harsh nuclear environments. To clarify the effect of irradiation on scale adhesion, alumina scales were pre-formed on FeCrAl ODS alloy via oxidation at 1000 °C and subsequently irradiated with 6.4 MeV Fe³⁺ ions at 500 °C at the Dual-Beam Facility for Energy Science and Technology, Kyoto University. Irradiation-induced segregation of reactive elements (REs), including Zr and Y, together with Ti/TiC precipitation at the grain-boundary triple junctions (GBTJs) of the alumina–alloy interface, was confirmed via scanning transmission electron microscopy. Atom probe tomography analysis further revealed irradiation-induced depletion of Ti, Zr, and Y, accompanied by Si impurity segregation at the oxide/alloy interface. The adhesion strength of the alumina scales before and after irradiation was quantitatively evaluated via micro double-notched shear compression tests, which indicated a slight increase in the interfacial adhesion strength after irradiation. This study suggests that the segregation of Ti/TiC and REs at the interfacial GBTJs under Fe-ion irradiation enhances the anchoring effect and improves the interfacial adhesion strength by promoting mechanical interlocking.
{"title":"Interfacial microstructure and adhesion evaluation of pre-oxidized alumina scales on FeCrAl ODS alloy under ion irradiation","authors":"Hao Yu , Diancheng Geng , Minha Park , Yasuyuki Ogino , Naoko Oono-Hori , Koji Inoue , Sosuke Kondo , Ryuta Kasada , Shigeharu Ukai","doi":"10.1016/j.jnucmat.2026.156464","DOIUrl":"10.1016/j.jnucmat.2026.156464","url":null,"abstract":"<div><div>Zr-added FeCrAl oxide-dispersion-strengthened (ODS) alloys have been developed as promising candidates for accident-tolerant fuel claddings, which can spontaneously form protective alumina scales at elevated temperatures and provide excellent oxidation resistance in harsh nuclear environments. To clarify the effect of irradiation on scale adhesion, alumina scales were pre-formed on FeCrAl ODS alloy via oxidation at 1000 °C and subsequently irradiated with 6.4 MeV Fe³⁺ ions at 500 °C at the Dual-Beam Facility for Energy Science and Technology, Kyoto University. Irradiation-induced segregation of reactive elements (REs), including Zr and Y, together with Ti/TiC precipitation at the grain-boundary triple junctions (GBTJs) of the alumina–alloy interface, was confirmed via scanning transmission electron microscopy. Atom probe tomography analysis further revealed irradiation-induced depletion of Ti, Zr, and Y, accompanied by Si impurity segregation at the oxide/alloy interface. The adhesion strength of the alumina scales before and after irradiation was quantitatively evaluated via micro double-notched shear compression tests, which indicated a slight increase in the interfacial adhesion strength after irradiation. This study suggests that the segregation of Ti/TiC and REs at the interfacial GBTJs under Fe-ion irradiation enhances the anchoring effect and improves the interfacial adhesion strength by promoting mechanical interlocking.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156464"},"PeriodicalIF":3.2,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024741","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}
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