Pub Date : 2025-12-29DOI: 10.1016/j.jnucmat.2025.156422
Haiyan Liao , Xiaohan Deng , Weijiu Huang , Haibo Ruan , Shuai Lyu , Yuan Niu , Xiangkong Xu , Yongyao Su , Junjun Wang
This study systematically investigates the effects of Al content (3, 5, and 7 wt%) and a Cr-N diffusion barrier on the high-temperature steam oxidation performance of magnetron-sputtered FeCrAl coatings deposited on Zry-4. The tests were conducted at 1200 °C under simulated loss of coolant accident (LOCA) conditions in a steam environment. Results reveal that a minimum Al content of 5 wt% is required to form a continuous and dense Al2O3 layer, which significantly improves oxidation resistance. The Cr-N interlayer effectively suppressed the interdiffusion of Zr and Fe by promoting the in-situ formation of a continuous Zr(Fe,Cr)2 Laves phase barrier during oxidation. This barrier inhibited the development of low-melting-point Fe-Zr eutectic phases, thereby enhancing interfacial stability. Compared with a pure Cr interlayer, the Cr-N barrier demonstrated superior performance in mitigating interdiffusion and improving coating durability. These findings highlight the critical roles of Al content and the Cr-N interlayer in optimizing FeCrAl coatings for accident-tolerant fuel (ATF) claddings, offering essential insights for designing zirconium alloy coatings with superior oxidation resistance and reliability under extreme conditions.
{"title":"Effect of Al content and a Cr-N diffusion barrier on the high-temperature steam oxidation of FeCrAl coatings on Zry-4","authors":"Haiyan Liao , Xiaohan Deng , Weijiu Huang , Haibo Ruan , Shuai Lyu , Yuan Niu , Xiangkong Xu , Yongyao Su , Junjun Wang","doi":"10.1016/j.jnucmat.2025.156422","DOIUrl":"10.1016/j.jnucmat.2025.156422","url":null,"abstract":"<div><div>This study systematically investigates the effects of Al content (3, 5, and 7 wt%) and a Cr-N diffusion barrier on the high-temperature steam oxidation performance of magnetron-sputtered FeCrAl coatings deposited on Zry-4. The tests were conducted at 1200 °C under simulated loss of coolant accident (LOCA) conditions in a steam environment. Results reveal that a minimum Al content of 5 wt% is required to form a continuous and dense Al<sub>2</sub>O<sub>3</sub> layer, which significantly improves oxidation resistance. The Cr-N interlayer effectively suppressed the interdiffusion of Zr and Fe by promoting the in-situ formation of a continuous Zr(Fe,Cr)<sub>2</sub> Laves phase barrier during oxidation. This barrier inhibited the development of low-melting-point Fe-Zr eutectic phases, thereby enhancing interfacial stability. Compared with a pure Cr interlayer, the Cr-N barrier demonstrated superior performance in mitigating interdiffusion and improving coating durability. These findings highlight the critical roles of Al content and the Cr-N interlayer in optimizing FeCrAl coatings for accident-tolerant fuel (ATF) claddings, offering essential insights for designing zirconium alloy coatings with superior oxidation resistance and reliability under extreme conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156422"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882269","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-12-28DOI: 10.1016/j.jnucmat.2025.156421
Jianguo Ma , Zhihong Liu , Chunwei Ma , Wei Wen , Huapeng Wu , Haibiao Ji , Rui Wang , Yuquan Kuang , Wangqi Shi , Haiying Xu , Weiping Fang , Zhiyong Wang , Yetao He
This paper systematically investigates the critical challenges in metal additive manufacturing (AM) technology for producing tungsten first wall components in nuclear fusion reactors—namely, residual stress and microcrack control. Research indicates that tungsten's high melting point (3422 °C) combined with AM's rapid cooling characteristics (10³-10⁴ K/s) results in melt pool temperature gradients exceeding 10⁶ K/m, inducing significant residual stresses. while the material exhibits a sharp decline in plasticity within the brittle-to-ductile transition temperature range of 450–650 K, making residual tensile stresses (>100 MPa) prone to triggering microcracks. Synergistic regulation through optimized scanning strategies (e.g., adjusting scan paths and energy distribution) and post-processing techniques achieves the following: Surface strengthening techniques (e.g., laser shock peening) form a 100–300 μm compressive stress layer, significantly reducing microcrack density; High-temperature heat treatment (1200–1400 °C) promotes crack healing through grain boundary migration. Data-driven machine learning methods enable real-time residual stress prediction, supporting intelligent process optimization. Future efforts should integrate multiscale simulations with irradiation experiments to validate and advance the engineering application of tungsten components in fusion reactors.
{"title":"Research progress on residual stress and microcrack control of tungsten fabricated via additive manufacturing","authors":"Jianguo Ma , Zhihong Liu , Chunwei Ma , Wei Wen , Huapeng Wu , Haibiao Ji , Rui Wang , Yuquan Kuang , Wangqi Shi , Haiying Xu , Weiping Fang , Zhiyong Wang , Yetao He","doi":"10.1016/j.jnucmat.2025.156421","DOIUrl":"10.1016/j.jnucmat.2025.156421","url":null,"abstract":"<div><div>This paper systematically investigates the critical challenges in metal additive manufacturing (AM) technology for producing tungsten first wall components in nuclear fusion reactors—namely, residual stress and microcrack control. Research indicates that tungsten's high melting point (3422 °C) combined with AM's rapid cooling characteristics (10³-10⁴ K/s) results in melt pool temperature gradients exceeding 10⁶ K/m, inducing significant residual stresses. while the material exhibits a sharp decline in plasticity within the brittle-to-ductile transition temperature range of 450–650 K, making residual tensile stresses (>100 MPa) prone to triggering microcracks. Synergistic regulation through optimized scanning strategies (e.g., adjusting scan paths and energy distribution) and post-processing techniques achieves the following: Surface strengthening techniques (e.g., laser shock peening) form a 100–300 μm compressive stress layer, significantly reducing microcrack density; High-temperature heat treatment (1200–1400 °C) promotes crack healing through grain boundary migration. Data-driven machine learning methods enable real-time residual stress prediction, supporting intelligent process optimization. Future efforts should integrate multiscale simulations with irradiation experiments to validate and advance the engineering application of tungsten components in fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156421"},"PeriodicalIF":3.2,"publicationDate":"2025-12-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922654","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 : 2025-12-28DOI: 10.1016/j.jnucmat.2025.156417
Chong Liu , Dazhao Cheng , Jiahui Qu , Dehui Li , Yan Zhao , Jing Zhang
The evolution of helium bubbles in nickel-based alloys poses significant challenges to the structural integrity of nuclear reactor components. This study investigates the spatial heterogeneity and anisotropic evolution of helium bubbles in Inconel 617 under He ion irradiation (1 MeV, 3 × 10¹⁶ ions/cm², 830 °C) using a coupled experimental and anisotropic phase-field approach. Transmission electron microscopy reveals distinct bubble characteristics: the (Ni,Cr)O oxide/matrix interface exhibits the highest bubble density, while both the interface and the oxide interior share similar, fine bubble sizes (∼2.4 nm). In contrast, the Ni matrix hosts larger bubbles (∼3.8 nm) at a lower density. The anisotropic phase-field model, developed to quantitatively interpret these observations, identifies the coherent oxide interface as a potent defect sink. Its strong absorption of helium and vacancies promotes dense bubble nucleation while inducing a local vacancy depletion that suppresses coarsening in the oxide. Conversely, in the matrix, long-range diffusion enables the growth of sparse, large bubbles. Critically, the simulation reproduces the experimentally observed quadrilateral bubble morphology in the matrix, confirming the dominance of surface energy anisotropy in shaping bubbles. This work provides the first quantitative interpretation of anisotropic bubble evolution at oxide interfaces in IN617, highlighting the critical interplay between defect sink strength and crystallographic anisotropy. The coupled methodology establishes a predictive framework for designing radiation-tolerant microstructures.
{"title":"Spatially heterogeneous evolution of helium bubbles in He-irradiated Inconel 617: Experimental observation and anisotropic phase-field simulation","authors":"Chong Liu , Dazhao Cheng , Jiahui Qu , Dehui Li , Yan Zhao , Jing Zhang","doi":"10.1016/j.jnucmat.2025.156417","DOIUrl":"10.1016/j.jnucmat.2025.156417","url":null,"abstract":"<div><div>The evolution of helium bubbles in nickel-based alloys poses significant challenges to the structural integrity of nuclear reactor components. This study investigates the spatial heterogeneity and anisotropic evolution of helium bubbles in Inconel 617 under He ion irradiation (1 MeV, 3 × 10¹⁶ ions/cm², 830 °C) using a coupled experimental and anisotropic phase-field approach. Transmission electron microscopy reveals distinct bubble characteristics: the (Ni,Cr)O oxide/matrix interface exhibits the highest bubble density, while both the interface and the oxide interior share similar, fine bubble sizes (∼2.4 nm). In contrast, the Ni matrix hosts larger bubbles (∼3.8 nm) at a lower density. The anisotropic phase-field model, developed to quantitatively interpret these observations, identifies the coherent oxide interface as a potent defect sink. Its strong absorption of helium and vacancies promotes dense bubble nucleation while inducing a local vacancy depletion that suppresses coarsening in the oxide. Conversely, in the matrix, long-range diffusion enables the growth of sparse, large bubbles. Critically, the simulation reproduces the experimentally observed quadrilateral bubble morphology in the matrix, confirming the dominance of surface energy anisotropy in shaping bubbles. This work provides the first quantitative interpretation of anisotropic bubble evolution at oxide interfaces in IN617, highlighting the critical interplay between defect sink strength and crystallographic anisotropy. The coupled methodology establishes a predictive framework for designing radiation-tolerant microstructures.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156417"},"PeriodicalIF":3.2,"publicationDate":"2025-12-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882273","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-12-27DOI: 10.1016/j.jnucmat.2025.156409
J.T. Rizk, X.-Y. Liu, D.A. Andersson, E. Kardoulaki, N.M. Abdul-Jabbar
The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. The DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.
{"title":"Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds","authors":"J.T. Rizk, X.-Y. Liu, D.A. Andersson, E. Kardoulaki, N.M. Abdul-Jabbar","doi":"10.1016/j.jnucmat.2025.156409","DOIUrl":"10.1016/j.jnucmat.2025.156409","url":null,"abstract":"<div><div>The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard <em>U</em> model (DFT+<em>U</em>) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+<em>U</em> calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. The DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156409"},"PeriodicalIF":3.2,"publicationDate":"2025-12-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922719","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156416
Calum S. Cunningham, Georgios Papanikos
Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT41J), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.
{"title":"Using machine learning to predict reactor pressure vessel embrittlement: Human factors and best practice","authors":"Calum S. Cunningham, Georgios Papanikos","doi":"10.1016/j.jnucmat.2025.156416","DOIUrl":"10.1016/j.jnucmat.2025.156416","url":null,"abstract":"<div><div>Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT<sub>41J</sub>), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156416"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957738","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156410
Ine Arts , Rolando Saniz , Gianguido Baldinozzi , Gregory Leinders , Marc Verwerft , Dirk Lamoen
The oxidation of UO2 is primarily governed by the diffusion of oxygen through the lattice. Oxygen diffusion is significantly influenced by defects and interfaces, with grain boundaries being particularly relevant in spent nuclear fuel due to their increasing concentration at the periphery of fuel rods. While experimental studies on bulk uranium suggest defects enhance oxygen diffusion, the role of grain boundaries remains contentious, with discrepancies between theoretical predictions of enhanced diffusion and experimental observations. This study employs density functional theory (DFT+U) to investigate the electronic properties and stability of interstitial oxygen in two coincident site lattice grain boundaries, Σ3 {111} and Σ5 {210}, in UO2. We compare stoichiometric and non-stoichiometric grain boundary models, examining their formation energies, defect interactions, and local structural distortions. The interstitial oxygen defects cause an expansion of the oxygen cage and a contraction of the U-O bonds, both in bulk and at the grain boundaries. The Σ3 grain boundary showed potential for defect accumulation, while the Σ5 grain boundary did not demonstrate reduced defect formation energies relative to the bulk. Our findings contribute to the understanding the UO2 oxidation process, in an attempt to address inconsistencies between theoretical and experimental studies on oxygen diffusion in grain boundaries.
{"title":"Electronic properties and stability of interstitial oxygen in UO2 grain boundaries: An ab initio study","authors":"Ine Arts , Rolando Saniz , Gianguido Baldinozzi , Gregory Leinders , Marc Verwerft , Dirk Lamoen","doi":"10.1016/j.jnucmat.2025.156410","DOIUrl":"10.1016/j.jnucmat.2025.156410","url":null,"abstract":"<div><div>The oxidation of UO<sub>2</sub> is primarily governed by the diffusion of oxygen through the lattice. Oxygen diffusion is significantly influenced by defects and interfaces, with grain boundaries being particularly relevant in spent nuclear fuel due to their increasing concentration at the periphery of fuel rods. While experimental studies on bulk uranium suggest defects enhance oxygen diffusion, the role of grain boundaries remains contentious, with discrepancies between theoretical predictions of enhanced diffusion and experimental observations. This study employs density functional theory (DFT+U) to investigate the electronic properties and stability of interstitial oxygen in two coincident site lattice grain boundaries, Σ3 {111} and Σ5 {210}, in UO<sub>2</sub>. We compare stoichiometric and non-stoichiometric grain boundary models, examining their formation energies, defect interactions, and local structural distortions. The interstitial oxygen defects cause an expansion of the oxygen cage and a contraction of the U-O bonds, both in bulk and at the grain boundaries. The Σ3 grain boundary showed potential for defect accumulation, while the Σ5 grain boundary did not demonstrate reduced defect formation energies relative to the bulk. Our findings contribute to the understanding the UO<sub>2</sub> oxidation process, in an attempt to address inconsistencies between theoretical and experimental studies on oxygen diffusion in grain boundaries.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156410"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882270","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156414
Canjia Huang , Zifeng Deng , Xingli Wang , Qiang Li , Wanjing Wang , Zhilu Liu , Jieyao He , Jiaxin Jin , Wei Cao , Zongxiao Guo , Fan Wang , Yunming Qiu , Ying Liu , Chunyan Yu , Shixing Wang , Jianjun Huang
The fabrication of high-quality W/steel first wall of blanket poses significant challenges in the field of nuclear fusion, primarily due to the substantial differences in thermodynamic properties between tungsten (W) and stainless steel (SS). Detonation sprayed (DS) iron (Fe) coating is a good alternative thermal stress-relieving interlayer between the DS-W armor and the steel substrate. This study adopted two post-heat treatment strategies to optimize the microstructure and improve the performance of the DS-Fe interlayer: carrying out vacuum heat treatments (ranging from 500 °C to 900 °C) on the DS-W/DS-Fe/SS structure and the DS-Fe/SS structure following the deposition of DS-W coating, respectively. The microstructures of the DS-Fe coatings and the DS-W/DS-Fe/SS structures after the heat treatments showed that three sets of these heat-treated specimens had better microstructure improvement than others. The electron beam-based transient high heat flux (HHF) loading tests revealed that the heat-treated specimens demonstrated notable improvements in the performance of the coating structure. Moreover, under transient HHF, the specimens subjected to 900 °C heat treatment after the deposition of DS-W coating exhibited remarkable performance, followed by those treated at 900 °C before the deposition of DS-W coating. The differences in the damage behaviors of the coating structures at 50 MJ/m² HHF between the two sets were studied in order to understand their performance characteristics and distinctions. These results confirmed that holistic heat treatment on the DS-W/DS-Fe/SS structure effectively alleviated the thermal stress in it under the fusion reactor conditions, thereby providing critical references for the manufacturing of the blanket first wall.
{"title":"Influence of heat treatment on the microstructure and performance of detonation sprayed W/Fe/steel first wall structure","authors":"Canjia Huang , Zifeng Deng , Xingli Wang , Qiang Li , Wanjing Wang , Zhilu Liu , Jieyao He , Jiaxin Jin , Wei Cao , Zongxiao Guo , Fan Wang , Yunming Qiu , Ying Liu , Chunyan Yu , Shixing Wang , Jianjun Huang","doi":"10.1016/j.jnucmat.2025.156414","DOIUrl":"10.1016/j.jnucmat.2025.156414","url":null,"abstract":"<div><div>The fabrication of high-quality W/steel first wall of blanket poses significant challenges in the field of nuclear fusion, primarily due to the substantial differences in thermodynamic properties between tungsten (W) and stainless steel (SS). Detonation sprayed (DS) iron (Fe) coating is a good alternative thermal stress-relieving interlayer between the DS-W armor and the steel substrate. This study adopted two post-heat treatment strategies to optimize the microstructure and improve the performance of the DS-Fe interlayer: carrying out vacuum heat treatments (ranging from 500 °C to 900 °C) on the DS-W/DS-Fe/SS structure and the DS-Fe/SS structure following the deposition of DS-W coating, respectively. The microstructures of the DS-Fe coatings and the DS-W/DS-Fe/SS structures after the heat treatments showed that three sets of these heat-treated specimens had better microstructure improvement than others. The electron beam-based transient high heat flux (HHF) loading tests revealed that the heat-treated specimens demonstrated notable improvements in the performance of the coating structure. Moreover, under transient HHF, the specimens subjected to 900 °C heat treatment after the deposition of DS-W coating exhibited remarkable performance, followed by those treated at 900 °C before the deposition of DS-W coating. The differences in the damage behaviors of the coating structures at 50 MJ/m² HHF between the two sets were studied in order to understand their performance characteristics and distinctions. These results confirmed that holistic heat treatment on the DS-W/DS-Fe/SS structure effectively alleviated the thermal stress in it under the fusion reactor conditions, thereby providing critical references for the manufacturing of the blanket first wall.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156414"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882338","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156415
L. Gubbels , J.P. Ramos , J. Vleugels , M. Verwerft , L. Popescu , B. Acevedo
In this study, both heterogeneous and homogeneous oxalate precipitation method, as well as hydroxide-carbonate precipitation, were employed to produce ThO₂ powders with distinct morphologies- platelets, cubes, and nanoparticles. These powders were compacted and sintered at 2023 K (1750 °C) under an oxygen potential of -420 kJ mol-1 for 1 h and 10 h. This approach enabled a systematic investigation of the influence of morphology on packing behavior, densification kinetics, and the resulting microstructure, which were characterized using XRD, SEM, BET surface area analysis, He pycnometry, and immersion density measurements. All powders achieved high final densities (>95 % of the theoretical density) after 10 h of sintering, underscoring their potential for nuclear fuel–related applications. For porous ThO₂ materials intended for low-temperature ISOL target applications, the oxalate-derived powders, particularly those precipitated at room temperature, provided a promising balance between low sinterability and process scalability.
{"title":"Sintering characteristics of ThO2 with different powder morphologies","authors":"L. Gubbels , J.P. Ramos , J. Vleugels , M. Verwerft , L. Popescu , B. Acevedo","doi":"10.1016/j.jnucmat.2025.156415","DOIUrl":"10.1016/j.jnucmat.2025.156415","url":null,"abstract":"<div><div>In this study, both heterogeneous and homogeneous oxalate precipitation method, as well as hydroxide-carbonate precipitation, were employed to produce ThO₂ powders with distinct morphologies- platelets, cubes, and nanoparticles. These powders were compacted and sintered at 2023 K (1750 °C) under an oxygen potential of -420 kJ mol<sup>-1</sup> for 1 h and 10 h. This approach enabled a systematic investigation of the influence of morphology on packing behavior, densification kinetics, and the resulting microstructure, which were characterized using XRD, SEM, BET surface area analysis, He pycnometry, and immersion density measurements. All powders achieved high final densities (>95 % of the theoretical density) after 10 h of sintering, underscoring their potential for nuclear fuel–related applications. For porous ThO₂ materials intended for low-temperature ISOL target applications, the oxalate-derived powders, particularly those precipitated at room temperature, provided a promising balance between low sinterability and process scalability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156415"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024747","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-12-25DOI: 10.1016/j.jnucmat.2025.156403
Tao Zhang , Qihang Liu , Yi-Lang Mai , Ya-Wen Li , Benxian Song , Hao Wang , Xiao-Chun Li , Fei Sun , Hai-Shan Zhou
The retention of tritium (T) in plasma-facing materials significantly impacts the operational lifespan and safety of fusion reactors. Hydrogen isotope (HI) exchange offers a viable strategy for T removal in such reactors. While hydrogen (H) clusters are known to influence HI exchange behavior, the specific mechanisms and patterns of their effects remain unclear. In this study, we employed extended nanosecond-scale molecular dynamics (MD) simulations to investigate the influence of temperature, H concentration, and vacancy concentration on the formation of H clusters and subsequent HI exchange in tungsten (W). Our results indicate that increasing the temperature consistently enhances the rate of T removal. As the H concentration rises, H evolves from isolated atoms into rod-shaped clusters along the {100} and {110} crystal plane families, and further expands into platelet-like structures. During the isolated atom stage, the T removal rate continues to increase. However, in the platelet-like cluster stage, the removal efficiency initially rises and then declines. Additionally, high vacancy concentrations promote the widening of the isolated H atom stage. These findings provide valuable insights for optimizing T removal parameters in fusion reactors.
{"title":"Molecular dynamics simulations of the influence of hydrogen clusters on hydrogen isotope exchange for tritium removal in tungsten vacancy","authors":"Tao Zhang , Qihang Liu , Yi-Lang Mai , Ya-Wen Li , Benxian Song , Hao Wang , Xiao-Chun Li , Fei Sun , Hai-Shan Zhou","doi":"10.1016/j.jnucmat.2025.156403","DOIUrl":"10.1016/j.jnucmat.2025.156403","url":null,"abstract":"<div><div>The retention of tritium (T) in plasma-facing materials significantly impacts the operational lifespan and safety of fusion reactors. Hydrogen isotope (HI) exchange offers a viable strategy for T removal in such reactors. While hydrogen (H) clusters are known to influence HI exchange behavior, the specific mechanisms and patterns of their effects remain unclear. In this study, we employed extended nanosecond-scale molecular dynamics (MD) simulations to investigate the influence of temperature, H concentration, and vacancy concentration on the formation of H clusters and subsequent HI exchange in tungsten (W). Our results indicate that increasing the temperature consistently enhances the rate of T removal. As the H concentration rises, H evolves from isolated atoms into rod-shaped clusters along the {100} and {110} crystal plane families, and further expands into platelet-like structures. During the isolated atom stage, the T removal rate continues to increase. However, in the platelet-like cluster stage, the removal efficiency initially rises and then declines. Additionally, high vacancy concentrations promote the widening of the isolated H atom stage. These findings provide valuable insights for optimizing T removal parameters in fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156403"},"PeriodicalIF":3.2,"publicationDate":"2025-12-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882266","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 : 2025-12-25DOI: 10.1016/j.jnucmat.2025.156413
YanBang Tang
Predicting the onset dose of void swelling is a critical challenge in developing radiation-resistant materials, a task often hindered by sparse and heterogeneous experimental data. To address this, we present a comprehensive framework combining a Gaussian noise-based data augmentation strategy with the AutoGluon automated machine learning (AutoML) platform. This study introduces a newly expanded dataset, updated with 80 recent publications (2020-2025) to form a comprehensive library of 374 irradiated metal samples. Our proposed framework's efficacy is rigorously evaluated by applying the augmentation strategy to AutoGluon, as well as to two state-of-the-art (SOTA) tabular models, TabM and TabPFN. The augmented AutoGluon model demonstrated superior performance, achieving a Root Mean Squared Error (RMSE) of 23.19 dpa and a coefficient of determination (R²) of 0.872 on an unseen test set. This represents a 6.3 % reduction in error compared to its baseline and outperforms the augmented SOTA models. The results consistently show that data augmentation improves performance across all model architectures. SHapley Additive exPlanations (SHAP) analysis of the superior model confirmed its physical interpretability, identifying key features and their complex interactions. This synergistic methodology demonstrates a powerful, validated pathway to overcome data scarcity in materials informatics, enhancing predictive power and accelerating the data-driven design of advanced alloys.
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