Journal of Nuclear Materials最新文献

英文中文

A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155562

M. Vergani , M.W.D. Cooper , L. Noirot

In the literature, a clear definition of the irradiation re-solution frequency of gas from bubbles in the UO₂ fuel is absent. Moreover, for intra-granular bubbles, a detailed calculation of the cumulated displaced gas quantities in function of the distance from the radius of the bubble after a re-solution event has never been published. The assessment of these two elements is very useful if we want to increase the adherence of fission gas release codes to our present knowledge of the behavior of fission gases. Hence, we suggest to link the definition of the re-solution frequency to atomic-scale simulations. Furthermore, we present the cumulated displaced gas quantities obtained from Molecular Dynamics calculations, from which we have derived a re-solution profile that can be exploited to better consider the irradiation re-solution phenomenon inside Fission Gas Release codes. On top of that, we have built a new trapping/re-solution model for intra-granular bubbles linked to Molecular Dynamics simulations that can be easily incorporated into Fission Gas Release codes. We also check that the model is properly built through the comparison of the new model against a reference.

{"title":"A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations","authors":"M. Vergani , M.W.D. Cooper , L. Noirot","doi":"10.1016/j.jnucmat.2024.155562","DOIUrl":"10.1016/j.jnucmat.2024.155562","url":null,"abstract":"<div><div>In the literature, a clear definition of the irradiation re-solution frequency of gas from bubbles in the UO<sub>2</sub> fuel is absent. Moreover, for intra-granular bubbles, a detailed calculation of the cumulated displaced gas quantities in function of the distance from the radius of the bubble after a re-solution event has never been published. The assessment of these two elements is very useful if we want to increase the adherence of fission gas release codes to our present knowledge of the behavior of fission gases. Hence, we suggest to link the definition of the re-solution frequency to atomic-scale simulations. Furthermore, we present the cumulated displaced gas quantities obtained from Molecular Dynamics calculations, from which we have derived a re-solution profile that can be exploited to better consider the irradiation re-solution phenomenon inside Fission Gas Release codes. On top of that, we have built a new trapping/re-solution model for intra-granular bubbles linked to Molecular Dynamics simulations that can be easily incorporated into Fission Gas Release codes. We also check that the model is properly built through the comparison of the new model against a reference.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155562"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171603","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development of irradiation- and high-temperature resistant steels for fusion applications: Belgian contribution

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155611

D. Terentyev , O. Kachko , A. Puype , S. Valiyev , K. Iroc , A. Zinovev

In this work, we investigate alternative routes for the production of reduced activation ferritic-martensitic (RAFM) steels aiming to achieve specific improvements of their performance under fusion operational conditions. The latter impose at least two specific challenges: (i) low-temperature embrittlement (LTE) and (ii) high-temperature creep (HTC) deformation. In this work, we review the optimization routes attempted to alleviate the above noted challenges which are otherwise met in EUROFER97 steel. The development routes include: (i) reduction of manganese and carbon content coupled with alternation of other chemical elements and followed by quench & rolling procedures; (ii) alternation of spatial distribution and structural morphology of carbonitrides by varying carbon, vanadium and tantalum content based on thermodynamic computations and followed by thermo-mechanical treatment optimization; (iii) doping with zirconium/titanium and increase of tantalum content to improve ductility and toughness. The targeted enhanced performance is achieved without compromising strength and DBTT. The results of the baseline characterization including mechanical tests and microstructural characterization are presented. The contribution of the microstructural features constituting the ferritic martensitic steels into the tensile strength is analyzed based on existing mechanistic models and discussed to rationalize the improvements achieved.

{"title":"Development of irradiation- and high-temperature resistant steels for fusion applications: Belgian contribution","authors":"D. Terentyev , O. Kachko , A. Puype , S. Valiyev , K. Iroc , A. Zinovev","doi":"10.1016/j.jnucmat.2025.155611","DOIUrl":"10.1016/j.jnucmat.2025.155611","url":null,"abstract":"<div><div>In this work, we investigate alternative routes for the production of reduced activation ferritic-martensitic (RAFM) steels aiming to achieve specific improvements of their performance under fusion operational conditions. The latter impose at least two specific challenges: (i) low-temperature embrittlement (LTE) and (ii) high-temperature creep (HTC) deformation. In this work, we review the optimization routes attempted to alleviate the above noted challenges which are otherwise met in EUROFER97 steel. The development routes include: (i) reduction of manganese and carbon content coupled with alternation of other chemical elements and followed by quench & rolling procedures; (ii) alternation of spatial distribution and structural morphology of carbonitrides by varying carbon, vanadium and tantalum content based on thermodynamic computations and followed by thermo-mechanical treatment optimization; (iii) doping with zirconium/titanium and increase of tantalum content to improve ductility and toughness. The targeted enhanced performance is achieved without compromising strength and DBTT. The results of the baseline characterization including mechanical tests and microstructural characterization are presented. The contribution of the microstructural features constituting the ferritic martensitic steels into the tensile strength is analyzed based on existing mechanistic models and discussed to rationalize the improvements achieved.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155611"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154915","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Microstructure evolution in titanium carbide with different stoichiometry under 3 MeV Au2+ ion irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155609

Jinyu Shi , Yiming Lei , Chenxu Wang , Jie Zhang , Jingyang Wang

Titanium carbide (TiC) with the merits of stability and corrosion resistance has been regarded as promising structural material candidate for advanced nuclear reactors. The effects of deviation in carbon stoichiometry and local ordering of carbon vacancies on the irradiation-induced microstructure evolution of TiC_x (x = 0.62–0.98) were targeted. 3 MeV Au²⁺ ion irradiation at room temperature (RT) was conducted over a series of ion fluences ranging from 1 × 10¹⁴ to 2 × 10¹⁶ ions cm^-2, together with grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM). No amorphization was traced for titanium carbide ceramics with different stoichiometry irradiated at doses up to ∼70 displacements per atom (dpa). Substoichiometric titanium carbides exhibited excellent lattice expansion resistance compared to near stoichiometric one beyond a dose of ∼30 dpa. In addition, irradiation-induced two ordered phases and twins were observed. Local ordering of C vacancies benefits the accommodation, annihilation of irradiation induced defects, which enhances the tolerance of irradiation-induced amorphization of titanium carbide ceramics. This work provides a comprehensive understanding of microstructure evolution in titanium carbide with different stoichiometry, which facilitates the application of titanium carbide ceramics as advanced reactors cores concepts.

{"title":"Microstructure evolution in titanium carbide with different stoichiometry under 3 MeV Au2+ ion irradiation","authors":"Jinyu Shi , Yiming Lei , Chenxu Wang , Jie Zhang , Jingyang Wang","doi":"10.1016/j.jnucmat.2025.155609","DOIUrl":"10.1016/j.jnucmat.2025.155609","url":null,"abstract":"<div><div>Titanium carbide (TiC) with the merits of stability and corrosion resistance has been regarded as promising structural material candidate for advanced nuclear reactors. The effects of deviation in carbon stoichiometry and local ordering of carbon vacancies on the irradiation-induced microstructure evolution of TiC<sub>x</sub> (<em>x</em> = 0.62–0.98) were targeted. 3 MeV Au<sup>2+</sup> ion irradiation at room temperature (RT) was conducted over a series of ion fluences ranging from 1 × 10<sup>14</sup> to 2 × 10<sup>16</sup> ions cm<sup>-2</sup>, together with grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM). No amorphization was traced for titanium carbide ceramics with different stoichiometry irradiated at doses up to ∼70 displacements per atom (dpa). Substoichiometric titanium carbides exhibited excellent lattice expansion resistance compared to near stoichiometric one beyond a dose of ∼30 dpa. In addition, irradiation-induced two ordered phases and twins were observed. Local ordering of C vacancies benefits the accommodation, annihilation of irradiation induced defects, which enhances the tolerance of irradiation-induced amorphization of titanium carbide ceramics. This work provides a comprehensive understanding of microstructure evolution in titanium carbide with different stoichiometry, which facilitates the application of titanium carbide ceramics as advanced reactors cores concepts.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155609"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154924","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

High-temperature oxidation of accident tolerant Cr-coated Zr alloy cladding: Model development and validation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155621

Dong Wang , Shihao Wu , Kai Lu , Yapei Zhang , G.H. Su , Xi Liu

In our previous studies of Cr-coated Zr alloy cladding, models for the oxidation and degradation of Cr coating and for the oxidation and phase transformation of Zr alloy have been established. On these foundations, oxygen flux models for Cr-Zr non-eutectic and eutectic conditions are developed in this paper, enabling a comprehensive simulation of the oxidation processes of Cr-coated Zr alloy. The models are validated through oxidation experiments under isothermal and temperature ramp conditions, showing good agreement with the experimental data. The models can effectively simulate the processes where, under non-eutectic conditions, oxygen atoms first dissolve into the prior β-Zr matrix after coating degradation, followed by the formation of α-Zr(O) and eventually ZrO₂. For conditions with eutectic reaction, model assumption of immediate complete coating failure could lead to overestimation of substrate oxidation extent at low heating rates. This prediction deviation decreases as heating rate increases.

引用次数: 0

Effect of trace Al and Ti elements on borosilicate glass corrosion resistance of Inconel 690 alloy

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155626

Jing Ma , Wenguo Xu , Ce Zheng , Yingju Li , Xiaohui Feng , Yuansheng Yang

The corrosion behavior of Inconel 690 alloys with trace Al and Ti elements was investigated in borosilicate glass at 1150 °C. The results showed that the dimensional losses and average corrosion rates were decreased with trace Al addition, which showed excellent glass corrosion resistance, with an annual average corrosion rate below 0.41 mm/a. Differently, the co-addition of Al and Ti increased the dimensional losses and average corrosion rates from 7 days to 21 days, but decreased at 42 days. The trace Al elements lead to the formation of Al₂O₃ oxides at the grain boundaries, which hinders the diffusion of Cr³⁺ along the grain boundaries. The slow diffusion of Cr³⁺ inhibits the growth of the Cr₂O₃ oxide film and reduces the number of holes in the substrate, leading to excellent glass corrosion resistance. With the co-addition of Al and Ti, the Ti element enhanced the Cr³⁺ diffusion from inner to outer and the formation of the inner oxide, resulting in the holes and continuous Cr₂O₃ particles along the grain boundaries. The substrates near the surfaces were peeled along grain boundaries, leading to worse glass corrosion resistance.

{"title":"Effect of trace Al and Ti elements on borosilicate glass corrosion resistance of Inconel 690 alloy","authors":"Jing Ma , Wenguo Xu , Ce Zheng , Yingju Li , Xiaohui Feng , Yuansheng Yang","doi":"10.1016/j.jnucmat.2025.155626","DOIUrl":"10.1016/j.jnucmat.2025.155626","url":null,"abstract":"<div><div>The corrosion behavior of Inconel 690 alloys with trace Al and Ti elements was investigated in borosilicate glass at 1150 °C. The results showed that the dimensional losses and average corrosion rates were decreased with trace Al addition, which showed excellent glass corrosion resistance, with an annual average corrosion rate below 0.41 mm/a. Differently, the co-addition of Al and Ti increased the dimensional losses and average corrosion rates from 7 days to 21 days, but decreased at 42 days. The trace Al elements lead to the formation of Al<sub>2</sub>O<sub>3</sub> oxides at the grain boundaries, which hinders the diffusion of Cr<sup>3+</sup> along the grain boundaries. The slow diffusion of Cr<sup>3+</sup> inhibits the growth of the Cr<sub>2</sub>O<sub>3</sub> oxide film and reduces the number of holes in the substrate, leading to excellent glass corrosion resistance. With the co-addition of Al and Ti, the Ti element enhanced the Cr<sup>3+</sup> diffusion from inner to outer and the formation of the inner oxide, resulting in the holes and continuous Cr<sub>2</sub>O<sub>3</sub> particles along the grain boundaries. The substrates near the surfaces were peeled along grain boundaries, leading to worse glass corrosion resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155626"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155320","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Retention, sputtering and surface chemistry at tungsten oxide surface facing deuterium plasma

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155622

Meral Sharkass , Swarit Dwivedi , Yun Kyung Shin , Martin Nieto-Perez , Adri C.T. van Duin , Predrag S. Krstic

Our study investigates the response of an oxidized tungsten surface to deuterium irradiation in the 5–120 eV impact energy range. Using the LAMMPS molecular dynamics tool and a ReaxFF force field, we analyze the retention, reflection, sputtering, and surface chemistry of oxidized layers at various thicknesses at room temperature. These layers, formed on the tungsten (001) surface through cumulative oxygen irradiation, show that most reflected D atoms and sputtered O atoms originate in the oxide ad-layer, not reaching the W-bulk. The retention probability of D atoms is notably high at the lowest energies, decreasing with energy and approaching metallic tungsten values at higher energies. Our findings, which are compared with those of metallic tungsten and existing literature, provide valuable insights into the behavior of oxidized tungsten surfaces under deuterium plasma irradiation, with potential applications in the design of plasma-facing components for fusion reactors.

引用次数: 0

Characterization of the microstructure of yttrium hydride under proton irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155586

Stephen Taller , Fabian Naab , Takaaki Koyanagi , Timothy Lach

High moderation per unit volume solid moderator materials like yttrium hydride (YH_x) are necessary for compact nuclear microreactors. However, the phase stability and hydrogen transport processes of YH_x under high-temperature irradiation are largely unknown. Proton irradiation was conducted on YH_x at 300 °C and 580 °C to 0.2 dpa using 1 MeV or 2 MeV protons in a high-vacuum environment. The hydrogen concentration was determined before and after irradiation using elastic recoil detection analysis, and microstructural evolution was examined via post-irradiation scanning transmission electron microscopy and Raman spectroscopy. Dislocation loops and cavities were observed in all conditions; their distribution was correlated with the bombarding proton energy and ion irradiation temperature. This work revealed that hydrogen retention is proportional to the formation of traps for hydrogen gas atoms and identified pathways for hydrogen release. The relative contributions of bulk or fast diffusion paths, such as grain boundaries, delamination boundaries, and stacking faults are discussed; the primary mechanisms of hydrogen loss are likely based on diffusion, ruling out artefacts of the experimental design. The study suggests proton irradiation may be a strong surrogate to study hydrogen transport in hydride moderator materials under irradiation.

{"title":"Characterization of the microstructure of yttrium hydride under proton irradiation","authors":"Stephen Taller , Fabian Naab , Takaaki Koyanagi , Timothy Lach","doi":"10.1016/j.jnucmat.2024.155586","DOIUrl":"10.1016/j.jnucmat.2024.155586","url":null,"abstract":"<div><div>High moderation per unit volume solid moderator materials like yttrium hydride (YH<sub>x</sub>) are necessary for compact nuclear microreactors. However, the phase stability and hydrogen transport processes of YH<em><sub>x</sub></em> under high-temperature irradiation are largely unknown. Proton irradiation was conducted on YH<em><sub>x</sub></em> at 300 °C and 580 °C to 0.2 dpa using 1 MeV or 2 MeV protons in a high-vacuum environment. The hydrogen concentration was determined before and after irradiation using elastic recoil detection analysis, and microstructural evolution was examined via post-irradiation scanning transmission electron microscopy and Raman spectroscopy. Dislocation loops and cavities were observed in all conditions; their distribution was correlated with the bombarding proton energy and ion irradiation temperature. This work revealed that hydrogen retention is proportional to the formation of traps for hydrogen gas atoms and identified pathways for hydrogen release. The relative contributions of bulk or fast diffusion paths, such as grain boundaries, delamination boundaries, and stacking faults are discussed; the primary mechanisms of hydrogen loss are likely based on diffusion, ruling out artefacts of the experimental design. The study suggests proton irradiation may be a strong surrogate to study hydrogen transport in hydride moderator materials under irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155586"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154923","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155663

Hao Yu , Diancheng Geng , Yasuyuki Ogino , Naoko Oono-Hori , Koji Inoue , Sosuke Kondo , Ryuta Kasada , Shigeharu Ukai

The design of FeCrAl ferritic oxide dispersion strengthened (ODS) alloys is based on the formation of a stable alumina scale, which is expected to protect the alloys from extreme heat and corrosion in nuclear applications. To ensure reliable alumina protection in nuclear environment, it is indispensable to concern the radiation tolerance of the alumina scales formed on the FeCrAl ODS alloys. The present study investigates the effect of Fe ions irradiation on fracture modes and grain boundary cohesion of the alumina scales in conjunction with nano-impact tests and micro-double notch shear (DNS) compression tests. Pre-oxidation was carried out in air at 1000 °C to form an α-alumina layer on the surface of Fe-15Cr-7Al-0.5Y₂O₃–0.4Zr (wt.%) ferritic ODS alloy, followed by 6.4 MeV Fe³⁺ ion beam irradiation at 500 °C. Based on the microstructural characterization of the cross-sectional micrographs of nanoindentation imprints on the alumina scales, it was confirmed that the irradiation on the alumina scales resulted in significant intergranular fracture in nanoindentation, whereas the unirradiated alumina scales showed transgranular fracture. The elemental distribution around the alumina grain boundaries was elucidated with the aid of scanning transmission electron microscopy (STEM) and atom probe tomography (APT) observations, and obvious segregation of reactive elements (REs) and intergranular Ti/TiC precipitation were observed after irradiation, indicating the link between the microstructural evolution and the fracture behavior of the alumina scales. The detailed grain boundary cohesion of alumina scales before and after irradiation was accurately measured by the micro-DNS compression tests, and the results showed that the cohesion strength of the alumina decreased significantly after the Fe ions irradiation.

{"title":"Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy","authors":"Hao Yu , Diancheng Geng , Yasuyuki Ogino , Naoko Oono-Hori , Koji Inoue , Sosuke Kondo , Ryuta Kasada , Shigeharu Ukai","doi":"10.1016/j.jnucmat.2025.155663","DOIUrl":"10.1016/j.jnucmat.2025.155663","url":null,"abstract":"<div><div>The design of FeCrAl ferritic oxide dispersion strengthened (ODS) alloys is based on the formation of a stable alumina scale, which is expected to protect the alloys from extreme heat and corrosion in nuclear applications. To ensure reliable alumina protection in nuclear environment, it is indispensable to concern the radiation tolerance of the alumina scales formed on the FeCrAl ODS alloys. The present study investigates the effect of Fe ions irradiation on fracture modes and grain boundary cohesion of the alumina scales in conjunction with nano-impact tests and micro-double notch shear (DNS) compression tests. Pre-oxidation was carried out in air at 1000 °C to form an α-alumina layer on the surface of Fe-15Cr-7Al-0.5Y<sub>2</sub>O<sub>3</sub>–0.4Zr (wt.%) ferritic ODS alloy, followed by 6.4 MeV Fe<sup>3+</sup> ion beam irradiation at 500 °C. Based on the microstructural characterization of the cross-sectional micrographs of nanoindentation imprints on the alumina scales, it was confirmed that the irradiation on the alumina scales resulted in significant intergranular fracture in nanoindentation, whereas the unirradiated alumina scales showed transgranular fracture. The elemental distribution around the alumina grain boundaries was elucidated with the aid of scanning transmission electron microscopy (STEM) and atom probe tomography (APT) observations, and obvious segregation of reactive elements (REs) and intergranular Ti/TiC precipitation were observed after irradiation, indicating the link between the microstructural evolution and the fracture behavior of the alumina scales. The detailed grain boundary cohesion of alumina scales before and after irradiation was accurately measured by the micro-DNS compression tests, and the results showed that the cohesion strength of the alumina decreased significantly after the Fe ions irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155663"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155303","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Effects of He+ energy and W temperature on the initial W fuzz growth under the fusion-relevant He+ irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155630

Weifeng Liu, Chunjie Niu, Weiyuan Ni, Dongping Liu

The growth of tungsten nanofuzz (W fuzz) induced by helium ions (He⁺) irradiation is a critical issue for fusion devices such as ITER. In our study, we investigated the behavior of tungsten (W) under helium ion (He⁺) irradiation, focusing on the conditions that promote the growth of tungsten fuzz (W fuzz). A comprehensive model was utilized to analyze the impact of varying He⁺ energies and W temperatures on the W fuzz formation and growth. It was found that W fuzz was formed in a low-energy range from ∼10 eV to 200 eV and a higher-energy range of >5 keV. W fuzz growth was facilitated below the He-W sputtering threshold energy while sputtering erosion became a significant hindrance just above this threshold. At He⁺ energies reaching hundreds of eV, W fuzz growth was entirely suppressed due to the enhanced sputtering erosion. However, at higher He⁺ energies around 5 keV, the He⁺ penetration reduced the impact of sputtering erosion, allowing He bubbles to grow in the deeper W layer. W temperature varying in the range of 1000 – 2000 K played a crucial role in forming He bubbles – induced tensile stress in the W surface layer, therefore affecting W fuzz growth which strongly depended on the He⁺ energy. These insights provided a detailed understanding of the energy-temperature relevance for W fuzz formation, which was crucial for predicting the performance and lifetime of W components in future fusion reactors.

{"title":"Effects of He+ energy and W temperature on the initial W fuzz growth under the fusion-relevant He+ irradiation","authors":"Weifeng Liu, Chunjie Niu, Weiyuan Ni, Dongping Liu","doi":"10.1016/j.jnucmat.2025.155630","DOIUrl":"10.1016/j.jnucmat.2025.155630","url":null,"abstract":"<div><div>The growth of tungsten nanofuzz (W fuzz) induced by helium ions (He<sup>+</sup>) irradiation is a critical issue for fusion devices such as ITER. In our study, we investigated the behavior of tungsten (W) under helium ion (He<sup>+</sup>) irradiation, focusing on the conditions that promote the growth of tungsten fuzz (W fuzz). A comprehensive model was utilized to analyze the impact of varying He<sup>+</sup> energies and W temperatures on the W fuzz formation and growth. It was found that W fuzz was formed in a low-energy range from ∼10 eV to 200 eV and a higher-energy range of >5 keV. W fuzz growth was facilitated below the He-W sputtering threshold energy while sputtering erosion became a significant hindrance just above this threshold. At He<sup>+</sup> energies reaching hundreds of eV, W fuzz growth was entirely suppressed due to the enhanced sputtering erosion. However, at higher He<sup>+</sup> energies around 5 keV, the He<sup>+</sup> penetration reduced the impact of sputtering erosion, allowing He bubbles to grow in the deeper W layer. W temperature varying in the range of 1000 – 2000 K played a crucial role in forming He bubbles – induced tensile stress in the W surface layer, therefore affecting W fuzz growth which strongly depended on the He<sup>+</sup> energy. These insights provided a detailed understanding of the energy-temperature relevance for W fuzz formation, which was crucial for predicting the performance and lifetime of W components in future fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155630"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155344","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The role of microstructural evolution in irradiation hardening of Alloy 718 under low dose proton irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155644

Haipeng Zhu , Zhiming Zhang , Jianqiu Wang , Hongliang Ming , Zhiyuan Zhang , Yilan Jiang , Quanyao Ren , En-Hou Han

The microstructural evolution and hardness changes of Alloy 718 under low dose (≤0.09 dpa) proton irradiation were investigated. Low dose irradiation induced the formation of defect clusters and disordering of the γ″ phase, while no voids or radiation-induced segregation were observed. The density of defect clusters increased with increasing irradiation dose, but their size did not change significantly. The irradiation damage behavior varied among different γ″ variants, with those having their [001] direction parallel to the proton beam being the most resistant to disordering. The hardness of the irradiated material was simultaneously affected by the hardening due to irradiation defects and the softening caused by the disordering of the γ″ phase. The competition between these two factors led to a decrease in hardness at 0.02 dpa and 0.04 dpa compared to the unirradiated material, while at 0.09 dpa, the hardness increased above that of the unirradiated material.

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