Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102005
Naoya Odaira , Katsuaki Kodama , Daisuke Ito , Yasushi Saito , Joseph Don Parker , Takenao Shinohara
Lead-bismuth eutectic has emerged as a promising candidate for liquid metal coolant for Gen-IV nuclear reactors. Lead-bismuth eutectic is a unique material that expands gradually within the solid state. It may induce pipe deformation or rupture if it solidifies in a pipe or a container. In this study, the strain distributions of lead–bismuth eutectic in stainless-steel cups were evaluated using wavelength-resolved neutron imaging method. The wettability-improved case exhibited significantly larger compressive strain than in the others. The adhesion between lead–bismuth eutectic and the inner surface of the cup was a critical issue in the present study.
{"title":"Analysis of strain distribution of lead–bismuth eutectic inside a stainless steel cup by wavelength-resolved neutron imaging","authors":"Naoya Odaira , Katsuaki Kodama , Daisuke Ito , Yasushi Saito , Joseph Don Parker , Takenao Shinohara","doi":"10.1016/j.nme.2025.102005","DOIUrl":"10.1016/j.nme.2025.102005","url":null,"abstract":"<div><div>Lead-bismuth eutectic has emerged as a promising candidate for liquid metal coolant for Gen-IV nuclear reactors. Lead-bismuth eutectic is a unique material that expands gradually within the solid state. It may induce pipe deformation or rupture if it solidifies in a pipe or a container. In this study, the strain distributions of lead–bismuth eutectic in stainless-steel cups were evaluated using wavelength-resolved neutron imaging method. The wettability-improved case exhibited significantly larger compressive strain than in the others. The adhesion between lead–bismuth eutectic and the inner surface of the cup was a critical issue in the present study.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102005"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145362924","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102010
Jiaxuan Li , Yan Li , Junjie Gong , Xiaotong Chen , Yongxin Wang , Bing Liu
In this study, the adsorption and diffusion behaviors of three typical fission products Cs, Sr and I on the (111) surfaces of SiC, NbC and ZrC were studied by the first-principles calculations. Cs atom tended to be adsorbed on the H sites of SiC (111) and ZrC (111) surfaces, while on the T site of NbC (111) surface. H site of SiC (111) surface and F sites of NbC (111) and ZrC (111) surfaces were the most preferable adsorption sites for Sr. I atom was more likely to be adsorbed on the T site, H site and F site of SiC (111) surface, NbC (111) surface and ZrC (111) surface, respectively. All the above adsorption processes were evidenced to be chemisorption of both ionic and covalent features according to electron structure analysis. The smallest diffusion coefficients of Cs and Sr on these three surfaces at 1373 K were 1.155 × 10−8 m2·s−1 on ZrC (111) surface and 1.544 × 10−8 m2·s−1 on SiC (111) surface, respectively, indicating that ZrC has the best ability to prevent the surface diffusion of Cs, while SiC was the best for preventing Sr. All three carbide ceramic coating materials could effectively resist the surface diffusion of I since the diffusion coefficients of I on them were all relatively low, with the smallest diffusion coefficient of 1.130 × 10−8 m2·s−1 on NbC (111) surface.
本研究采用第一性原理计算方法研究了三种典型裂变产物Cs、Sr和I在SiC、NbC和ZrC（111）表面的吸附和扩散行为。Cs原子倾向于吸附在SiC（111）和ZrC（111）表面的H位，而Cs原子倾向于吸附在NbC（111）表面的T位。SiC（111）表面的H位、NbC（111）和ZrC（111）表面的F位是sr的最佳吸附位点，I原子更容易分别吸附在SiC（111）表面、NbC（111）表面和ZrC（111）表面的T位、H位和F位上。电子结构分析表明，上述吸附过程均为离子和共价的化学吸附。最小的扩散系数Cs和Sr这三个表面在1373 K 1.155×10−8平方米·s−1 ZrC表面(111)和1.544×10−8平方米·s−1 SiC(111)表面,分别表明ZrC有最好的防止Cs的表面扩散的能力,而碳化硅是最好的防止老三个碳化物陶瓷涂层材料能有效地抵抗的表面扩散我自扩散系数都相对较低,在NbC（111）表面的扩散系数最小，为1.130 × 10−8 m2·s−1。
{"title":"First principles investigation of adsorption and diffusion behavior of fission products (Cs, Sr and I) on carbide ceramic (SiC, NbC and ZrC) coatings","authors":"Jiaxuan Li , Yan Li , Junjie Gong , Xiaotong Chen , Yongxin Wang , Bing Liu","doi":"10.1016/j.nme.2025.102010","DOIUrl":"10.1016/j.nme.2025.102010","url":null,"abstract":"<div><div>In this study, the adsorption and diffusion behaviors of three typical fission products Cs, Sr and I on the (111) surfaces of SiC, NbC and ZrC were studied by the first-principles calculations. Cs atom tended to be adsorbed on the H sites of SiC (111) and ZrC (111) surfaces, while on the T site of NbC (111) surface. H site of SiC (111) surface and F sites of NbC (111) and ZrC (111) surfaces were the most preferable adsorption sites for Sr. I atom was more likely to be adsorbed on the T site, H site and F site of SiC (111) surface, NbC (111) surface and ZrC (111) surface, respectively. All the above adsorption processes were evidenced to be chemisorption of both ionic and covalent features according to electron structure analysis. The smallest diffusion coefficients of Cs and Sr on these three surfaces at 1373 K were 1.155 × 10<sup>−8</sup> m<sup>2</sup>·s<sup>−1</sup> on ZrC (111) surface and 1.544 × 10<sup>−8</sup> m<sup>2</sup>·s<sup>−1</sup> on SiC (111) surface, respectively, indicating that ZrC has the best ability to prevent the surface diffusion of Cs, while SiC was the best for preventing Sr. All three carbide ceramic coating materials could effectively resist the surface diffusion of I since the diffusion coefficients of I on them were all relatively low, with the smallest diffusion coefficient of 1.130 × 10<sup>−8</sup> m<sup>2</sup>·s<sup>−1</sup> on NbC (111) surface.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102010"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145363703","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102008
Zeyuan Sun , Xijie Wu , Ao Liu , Jie Pan , Zixie Wang , Mengli Li , Qiliang Mei , Jun Li , Xueshan Xiao
The phase transition behaviour and properties of Ti-Gd based alloys were investigated. The addition of Gd increases the α/β phase transition temperature of titanium alloys. The equiaxial α-phase in Ti-xGd alloys transforms into lamellar α-phase from 800 °C to 1000 °C. The addition of Fe element reduces the α/β phase transition temperature, and the α + β → β transition in Ti-7.5Fe-xGd alloys at temperatures between 600 °C and 800 °C. Gd elements precipitated as a second phase and alloys slip were impeded by the Gd phase during plastic deformation. The Ti-5.0Gd alloy exhibited good mechanical properties, with tensile strengths exceeding 500 MPa and elongation reaching up to 19.2 %. In addition, as the gadolinium content increases, there is a significant decrease in thermal neutron transmittance. When the alloy thickness is 0.07 cm, the shielding efficiency of the Ti-10.0Gd alloy for neutrons with an energy of 0.025 eV approaches 100 %. This indicates that the material exhibits excellent mechanical properties and effective thermal neutron shielding capabilities.
{"title":"Phase transition behaviour and properties of Ti-Gd based thermal neutron shielding alloys","authors":"Zeyuan Sun , Xijie Wu , Ao Liu , Jie Pan , Zixie Wang , Mengli Li , Qiliang Mei , Jun Li , Xueshan Xiao","doi":"10.1016/j.nme.2025.102008","DOIUrl":"10.1016/j.nme.2025.102008","url":null,"abstract":"<div><div>The phase transition behaviour and properties of Ti-Gd based alloys were investigated. The addition of Gd increases the α/β phase transition temperature of titanium alloys. The equiaxial α-phase in Ti-<em>x</em>Gd alloys transforms into lamellar α-phase from 800 °C to 1000 °C. The addition of Fe element reduces the α/β phase transition temperature, and the α + β → β transition in Ti-7.5Fe-<em>x</em>Gd alloys at temperatures between 600 °C and 800 °C. Gd elements precipitated as a second phase and alloys slip were impeded by the Gd phase during plastic deformation. The Ti-5.0Gd alloy exhibited good mechanical properties, with tensile strengths exceeding 500 MPa and elongation reaching up to 19.2 %. In addition, as the gadolinium content increases, there is a significant decrease in thermal neutron transmittance. When the alloy thickness is 0.07 cm, the shielding efficiency of the Ti-10.0Gd alloy for neutrons with an energy of 0.025 eV approaches 100 %. This indicates that the material exhibits excellent mechanical properties and effective thermal neutron shielding capabilities.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102008"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321277","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102006
Seyit ÇAĞLAR
This study examines the impact of boron (B) reinforcement on the corrosion resistance, microstructural properties, mechanical performance, and radiation shielding capabilities of Al-10Sm2O3 composites. Comprehensive characterization studies were carried out by fabricating composites with 0B, 3B, 9B, 15B, 20B, and 30B contents. Density measurements revealed that increased B content increases macro and microporosity, decreasing the relative density. Hardness measurements showed a significant increase in hardness values attributed to the homogeneous distribution of B particles, with the highest hardness value recorded at 259.9 HV in the 30B composite. Wear tests indicate that increasing the B content enhances wear resistance and reduces material loss. Corrosion tests revealed an adverse change in corrosion potential and an increase in corrosion current density with increasing boron (B) content, indicating that B particles negatively affect corrosion resistance by disrupting the continuity of the oxide layer. Radiation shielding analyses performed using MCNP6.2 simulation showed that higher B content increases thermal and fast neutron macroscopic cross sections, whereas gamma-ray attenuation properties decrease. The findings indicate that B reinforcement improved the mechanical and tribological performance but reduced the corrosion resistance. However, the increased thermal and fast neutron macroscopic cross-sections reveal the potential of B-reinforced Al-10Sm2O3 composites for neutron shielding applications.
{"title":"Corrosion behavior, microstructure, and mechanical properties of Al-10Sm2O3-B neutron shielding composites","authors":"Seyit ÇAĞLAR","doi":"10.1016/j.nme.2025.102006","DOIUrl":"10.1016/j.nme.2025.102006","url":null,"abstract":"<div><div>This study examines the impact of boron (B) reinforcement on the corrosion resistance, microstructural properties, mechanical performance, and radiation shielding capabilities of Al-10Sm<sub>2</sub>O<sub>3</sub> composites. Comprehensive characterization studies were carried out by fabricating composites with 0B, 3B, 9B, 15B, 20B, and 30B contents. Density measurements revealed that increased B content increases macro and microporosity, decreasing the relative density. Hardness measurements showed a significant increase in hardness values attributed to the homogeneous distribution of B particles, with the highest hardness value recorded at 259.9 HV in the 30B composite. Wear tests indicate that increasing the B content enhances wear resistance and reduces material loss. Corrosion tests revealed an adverse change in corrosion potential and an increase in corrosion current density with increasing boron (B) content, indicating that B particles negatively affect corrosion resistance by disrupting the continuity of the oxide layer. Radiation shielding analyses performed using MCNP6.2 simulation showed that higher B content increases thermal and fast neutron macroscopic cross sections, whereas gamma-ray attenuation properties decrease. The findings indicate that B reinforcement improved the mechanical and tribological performance but reduced the corrosion resistance. However, the increased thermal and fast neutron macroscopic cross-sections reveal the potential of B-reinforced Al-10Sm<sub>2</sub>O<sub>3</sub> composites for neutron shielding applications.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102006"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321268","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102001
Qirong Zheng , Panfei Tang , Jun Liu , Chuanguo Zhang , Yonggang Li , Zhi Zeng
Understanding radiation damage and its relation to materials degradation is critical for the construction of safe and long-operating nuclear reactors. The mean-field cluster dynamics (CD) model predicts microstructure evolution over longer spatio-temporal scales and damage accumulation up to high doses, which is typically used to relate damage production to material properties; however, its inability to treat intra-cascade spatial correlations (SC) limits the accuracy of radiation damage modeling. Beyond the current solutions of using an ambiguous annealing time for cascade homogenization, we develop a novel characteristic-time annealing methodology that optimally embeds intra-cascade spatial correlations into the source term, establishing the CD-SC model within the continuum framework. This approach enables unified modeling of defect evolution under both monoenergetic ion bombardment and continuous neutron spectrum irradiation. CD-SC achieves kinetic Monte Carlo-level accuracy with high efficiency, reproducing experimental size distributions of dislocation loops/voids and dose trends under self-ion/neutron irradiations. The characteristic times for iron and tungsten are in the orders of μs and ps, respectively, breaking the traditionally accepted values (∼ns) under room-temperature conditions. We further reveal that the characteristic time is proportional to the relative positions of primary defects and inversely proportional to the competition between defect reaction and sink absorption, which are related to damage energy as well as temperature and material properties (sink character and migration energy), respectively. Finally, CD-SC is applied to predict neutron radiation damage under fission and fusion conditions. The present approach builds an accurate defect production term, enabling reliable evaluation of in-service materials and facilitating the development of next-generation materials.
{"title":"Optimally incorporating intra-cascade spatial correlations into mean-field cluster dynamics for radiation damage modeling","authors":"Qirong Zheng , Panfei Tang , Jun Liu , Chuanguo Zhang , Yonggang Li , Zhi Zeng","doi":"10.1016/j.nme.2025.102001","DOIUrl":"10.1016/j.nme.2025.102001","url":null,"abstract":"<div><div>Understanding radiation damage and its relation to materials degradation is critical for the construction of safe and long-operating nuclear reactors. The mean-field cluster dynamics (CD) model predicts microstructure evolution over longer spatio-temporal scales and damage accumulation up to high doses, which is typically used to relate damage production to material properties; however, its inability to treat intra-cascade spatial correlations (SC) limits the accuracy of radiation damage modeling. Beyond the current solutions of using an ambiguous annealing time for cascade homogenization, we develop a novel characteristic-time annealing methodology that optimally embeds intra-cascade spatial correlations into the source term, establishing the CD-SC model within the continuum framework. This approach enables unified modeling of defect evolution under both monoenergetic ion bombardment and continuous neutron spectrum irradiation. CD-SC achieves kinetic Monte Carlo-level accuracy with high efficiency, reproducing experimental size distributions of dislocation loops/voids and dose trends under self-ion/neutron irradiations. The characteristic times for iron and tungsten are in the orders of μs and ps, respectively, breaking the traditionally accepted values (∼ns) under room-temperature conditions. We further reveal that the characteristic time is proportional to the relative positions of primary defects and inversely proportional to the competition between defect reaction and sink absorption, which are related to damage energy as well as temperature and material properties (sink character and migration energy), respectively. Finally, CD-SC is applied to predict neutron radiation damage under fission and fusion conditions. The present approach builds an accurate defect production term, enabling reliable evaluation of in-service materials and facilitating the development of next-generation materials.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102001"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145528699","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-17DOI: 10.1016/j.nme.2025.102012
Yaozhi Li , Xinle Li , Qitao Wang , Mengjie Yin , Yanfen Li
To enhance the high-temperature performance and extend the upper service temperature of oxide dispersion strengthened (ODS) steels for fusion reactor applications, a novel 9Cr-ODS steel was designed by co-adding 0.2 wt% VN and 0.35 wt% Y2O3 powders during mechanical alloying. This synergistic design strategy aims to combine the benefits of MX-type nitrides and nano-oxides. For comparison, steels with individual additions of VN or Y2O3 were also fabricated. All three steels exhibited ultra-fine microstructures with submicron grain sizes, high dislocation densities, and a dense dispersion of nanoparticles. The steel with VN powder alone showed excellent ductility at 700 °C (∼40 % elongation), while the steel with Y2O3 powder alone achieved superior strength through effective dispersion strengthening. Remarkably, the co-addition of VN and Y2O3 powders led to the formation of V-N-rich Y2O3 composite nanoparticles with the smallest average size (∼6.4 nm) and highest number density (∼7.9 × 1022 m−3), resulting in a significant enhancement in tensile strength without compromising ductility. This strategy offers a promising route to tailor both strength and ductility in advanced ODS steels, providing a robust framework for designing nuclear structural materials with superior high-temperature performance and radiation resistance.
{"title":"Developing a novel ODS steel to tailor mechanical properties and microstructure by co-adding VN and Y2O3 powders","authors":"Yaozhi Li , Xinle Li , Qitao Wang , Mengjie Yin , Yanfen Li","doi":"10.1016/j.nme.2025.102012","DOIUrl":"10.1016/j.nme.2025.102012","url":null,"abstract":"<div><div>To enhance the high-temperature performance and extend the upper service temperature of oxide dispersion strengthened (ODS) steels for fusion reactor applications, a novel 9Cr-ODS steel was designed by co-adding 0.2 wt% VN and 0.35 wt% Y<sub>2</sub>O<sub>3</sub> powders during mechanical alloying. This synergistic design strategy aims to combine the benefits of MX-type nitrides and nano-oxides. For comparison, steels with individual additions of VN or Y<sub>2</sub>O<sub>3</sub> were also fabricated. All three steels exhibited ultra-fine microstructures with submicron grain sizes, high dislocation densities, and a dense dispersion of nanoparticles. The steel with VN powder alone showed excellent ductility at 700 °C (∼40 % elongation), while the steel with Y<sub>2</sub>O<sub>3</sub> powder alone achieved superior strength through effective dispersion strengthening. Remarkably, the co-addition of VN and Y<sub>2</sub>O<sub>3</sub> powders led to the formation of V-N-rich Y<sub>2</sub>O<sub>3</sub> composite nanoparticles with the smallest average size (∼6.4 nm) and highest number density (∼7.9 × 10<sup>22</sup> m<sup>−3</sup>), resulting in a significant enhancement in tensile strength without compromising ductility. This strategy offers a promising route to tailor both strength and ductility in advanced ODS steels, providing a robust framework for designing nuclear structural materials with superior high-temperature performance and radiation resistance.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102012"},"PeriodicalIF":2.7,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145362926","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-16DOI: 10.1016/j.nme.2025.102002
Z. Shen, T. Schwarz-Selinger, A. Manhard, M. Balden
A systematic investigation was carried out to study the effects of ion energy and flux during deuterium (D) exposure of self-ion damaged tungsten (W) at a sample temperature of 570 K. Experimental conditions included combinations of 5 and 38 eV/D ions and fluxes of and . The depth distribution of deuterium at the topmost 7.4 m was determined by He Nuclear Reaction Analysis (NRA), while its total inventory was evaluated using both NRA and Thermal Desorption Spectroscopy (TDS). Morphological modifications at the surface were analyzed by confocal laser scanning microscopy (CLSM) and scanning electron microscopy (SEM) together with focused ion beam cutting (FIB). The experimental results show that already a flux of leads to the formation of blisters on the W surface even for 5 eV/D. With 38 eV/D ions additional defects are created that trap deuterium exceeding a depth of 7.4 m. All blisters in this study were very shallow with widths in the micrometer range and heights of only few tens of nanometers. The study revealed that blisters with a low height-to-diameter ratio are difficult to detect using scanning electron microscopy (SEM). These features could only be clearly identified through CLSM applying the differential interference contrast (DIC) mode.
{"title":"Effect of ion energy and flux during deuterium plasma exposure of displacement-damaged tungsten","authors":"Z. Shen, T. Schwarz-Selinger, A. Manhard, M. Balden","doi":"10.1016/j.nme.2025.102002","DOIUrl":"10.1016/j.nme.2025.102002","url":null,"abstract":"<div><div>A systematic investigation was carried out to study the effects of ion energy and flux during deuterium (D) exposure of self-ion damaged tungsten (W) at a sample temperature of 570 K. Experimental conditions included combinations of 5 and 38 eV/D ions and fluxes of <span><math><mrow><mn>6</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>19</mn></mrow></msup></mrow></math></span> and <span><math><mrow><mn>5</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>20</mn></mrow></msup><mspace></mspace><mi>D</mi><mo>/</mo><msup><mrow><mi>m</mi></mrow><mrow><mn>2</mn></mrow></msup><mo>/</mo><mi>s</mi></mrow></math></span>. The depth distribution of deuterium at the topmost 7.4 <span><math><mi>μ</mi></math></span>m was determined by <span><math><msup><mrow></mrow><mrow><mn>3</mn></mrow></msup></math></span>He Nuclear Reaction Analysis (NRA), while its total inventory was evaluated using both NRA and Thermal Desorption Spectroscopy (TDS). Morphological modifications at the surface were analyzed by confocal laser scanning microscopy (CLSM) and scanning electron microscopy (SEM) together with focused ion beam cutting (FIB). The experimental results show that already a flux of <span><math><mrow><mn>5</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>20</mn></mrow></msup><mspace></mspace><mi>D</mi><mo>/</mo><msup><mrow><mi>m</mi></mrow><mrow><mn>2</mn></mrow></msup><mo>/</mo><mi>s</mi></mrow></math></span> leads to the formation of blisters on the W surface even for 5 eV/D. With 38 eV/D ions additional defects are created that trap deuterium exceeding a depth of 7.4 <span><math><mi>μ</mi></math></span>m. All blisters in this study were very shallow with widths in the micrometer range and heights of only few tens of nanometers. The study revealed that blisters with a low height-to-diameter ratio are difficult to detect using scanning electron microscopy (SEM). These features could only be clearly identified through CLSM applying the differential interference contrast (DIC) mode.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102002"},"PeriodicalIF":2.7,"publicationDate":"2025-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145362925","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-16DOI: 10.1016/j.nme.2025.102007
Chao Liu , Kongfang Wei , Tielong Shen , Peng Jin , Jing Li , Le Qi , Zhiguang Wang
The evolution of He bubbles during the oxidation process was investigated in He-implanted ferritic-martensitic steels exposed to high-temperature steam. Similar He bubble bands formed in the SIMP and T91 steels after 500 keV He2+ irradiation at 450 °C with a fluence of 1.0 × 1017 ions/cm2, and duplex oxide films composed porous (Fe, Cr)3O4 inner layer and dense Fe3O4 outer layer formed on the both steels after 450 °C steam oxidation for 100 h. He bubbles displayed obviously different evolution between the oxide layer not reaching the bubble band and the oxide layer exceeding the bubble band. The significant size increase and density decrease of the visible bubbles and outward shift of bubble band occurred in the latter case, while the bubbles size/density and depth distribution were almost unchanged in the former case, indicating that the He bubbles were largely affected by the local oxidation of the alloy matrix in the He bubbles distribution area. As the increase of He bubbles size not only has obvious effects on material properties such as embrittlement and swelling, but also causes the degradation of the protective performance of oxide films, the interaction effect of corrosion and He bubbles should be paid more attention in advanced nuclear systems such as the high He yield accelerator-driven system.
{"title":"Evolution of He bubbles during the steam oxidation in He-implanted ferritic-martensitic steels","authors":"Chao Liu , Kongfang Wei , Tielong Shen , Peng Jin , Jing Li , Le Qi , Zhiguang Wang","doi":"10.1016/j.nme.2025.102007","DOIUrl":"10.1016/j.nme.2025.102007","url":null,"abstract":"<div><div>The evolution of He bubbles during the oxidation process was investigated in He-implanted ferritic-martensitic steels exposed to high-temperature steam. Similar He bubble bands formed in the SIMP and T91 steels after 500 keV He<sup>2+</sup> irradiation at 450 °C with a fluence of 1.0 × 10<sup>17</sup> ions/cm<sup>2</sup>, and duplex oxide films composed porous (Fe, Cr)<sub>3</sub>O<sub>4</sub> inner layer and dense Fe<sub>3</sub>O<sub>4</sub> outer layer formed on the both steels after 450 °C steam oxidation for 100 h. He bubbles displayed obviously different evolution between the oxide layer not reaching the bubble band and the oxide layer exceeding the bubble band. The significant size increase and density decrease of the visible bubbles and outward shift of bubble band occurred in the latter case, while the bubbles size/density and depth distribution were almost unchanged in the former case, indicating that the He bubbles were largely affected by the local oxidation of the alloy matrix in the He bubbles distribution area. As the increase of He bubbles size not only has obvious effects on material properties such as embrittlement and swelling, but also causes the degradation of the protective performance of oxide films, the interaction effect of corrosion and He bubbles should be paid more attention in advanced nuclear systems such as the high He yield accelerator-driven system.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102007"},"PeriodicalIF":2.7,"publicationDate":"2025-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145363704","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-15DOI: 10.1016/j.nme.2025.102003
Marc Sackers , Oleksandr Marchuk , Anne Houben , Eduard Warkentin , Marcin Rasinski , Sebastijan Brezinsek , Arkadi Kreter
This study reports on deuterium plasma-boron layer interaction investigated at the linear plasma device PSI-2. In preparation, RF magnetron sputtering deposited 115 nm thick boron layers onto tungsten substrates. Exposing the samples to a deuterium plasma in PSI-2 provided the bombardment by ions. The discharge conditions in PSI-2 were chosen such that the bombardment was predominantly due to deuterons (D) impinging at normal incidence. This operating scenario yielded a typical ion flux density onto the sample of 4 1021 m−2 s−1. Varying the bias applied to the sample from -43 V (floating potential) to -100 V allowed for investigating the near-threshold erosion regime.
The B I 2p-3s transition (249.7 nm), and the BD A-X transition (432.8 nm) provided the time-resolved (10 s) spectroscopic fingerprint of physical and chemical erosion during the plasma discharges of 70 s duration. The emission of the BD molecular band depends only weakly on the bias applied to the sample. Notably, this emission is also present at the lowest impact energies, suggesting a contribution by chemical erosion. Contrary to that, atomic boron emission follows reasonably close to the expectations by the binary collision approximation (BCA). The post-mortem layer thickness characterization reveals net sputtering yields that exceed the predictions by BCA and agree well with data from Hechtl et al. (1992).
{"title":"Erosion of thin boron films at the linear plasma device PSI-2 during deuterium discharges: Atomic and molecular spectroscopy of boron","authors":"Marc Sackers , Oleksandr Marchuk , Anne Houben , Eduard Warkentin , Marcin Rasinski , Sebastijan Brezinsek , Arkadi Kreter","doi":"10.1016/j.nme.2025.102003","DOIUrl":"10.1016/j.nme.2025.102003","url":null,"abstract":"<div><div>This study reports on deuterium plasma-boron layer interaction investigated at the linear plasma device PSI-2. In preparation, RF magnetron sputtering deposited 115<!--> <!-->nm thick boron layers onto tungsten substrates. Exposing the samples to a deuterium plasma in PSI-2 provided the bombardment by ions. The discharge conditions in PSI-2 were chosen such that the bombardment was predominantly due to deuterons (D<span><math><msup><mrow></mrow><mrow><mo>+</mo></mrow></msup></math></span>) impinging at normal incidence. This operating scenario yielded a typical ion flux density onto the sample of 4 <span><math><mo>×</mo></math></span> 10<sup>21</sup> <!-->m<sup>−2</sup> <!-->s<sup>−1</sup>. Varying the bias applied to the sample from -43<!--> <!-->V (floating potential) to -100<!--> <!-->V allowed for investigating the near-threshold erosion regime.</div><div>The B I 2p-3s transition (249.7<!--> <!-->nm), and the BD A-X transition (432.8<!--> <!-->nm) provided the time-resolved (10<!--> <!-->s) spectroscopic fingerprint of physical and chemical erosion during the plasma discharges of 70 s duration. The emission of the BD molecular band depends only weakly on the bias applied to the sample. Notably, this emission is also present at the lowest impact energies, suggesting a contribution by chemical erosion. Contrary to that, atomic boron emission follows reasonably close to the expectations by the binary collision approximation (BCA). The post-mortem layer thickness characterization reveals net sputtering yields that exceed the predictions by BCA and agree well with data from Hechtl et al. (1992).</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102003"},"PeriodicalIF":2.7,"publicationDate":"2025-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321267","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-10-15DOI: 10.1016/j.nme.2025.102004
Indrek Jõgi , Peeter Paris , Marta Malin Muru , Mathilde Diez , Emmanuelle Tsitrone , Jari Likonen , Antti Hakola , Eduard Grigore , the WEST team
Laser Induced Breakdown Spectroscopy (LIBS) is a chemical analysis method that has been used in various remote handling applications and is currently being developed for the remote composition analysis of layers deposited on the first walls and divertor components of fusion reactors. The technique applies laser pulses to ablate a small amount of material, which forms a plasma plume and emits a spectrum characterizing the elements originating from the investigated material. The use of consecutive laser pulses at the same spot allows us to obtain elemental depth profiles with a resolution depending on the laser ablation rate. The aim of the present study was to investigate the feasibility of the LIBS method to analyze the boron (B) content in the surface layer of WEST divertor marker tiles removed after the 2019 experimental campaign (C4) when they were subjected to 16 boronizations, 3 during the C3 campaign and 13 during the C4 campaign. Boronizations are used in tungsten wall tokamaks to reduce the W contamination in core plasma, but the formed boron layers retain their effect only for a limited time due to erosion and redeposition. The six studied samples originated from different poloidal locations along the divertor including erosion and deposition-dominated regions. The LIBS spectra contained the emission lines of tungsten (W), molybdenum (Mo) and carbon (C), which were the main elements of the marker tile layers, and B as one of the species in the surface layers deposited during the experimental campaigns. The depth profiles showing the LIBS line intensities as a function of the applied laser pulse number at the same spot were generally consistent with the Glow Discharge Optical Emission Spectroscopy and Secondary Ion Mass Spectrometry depth profiles obtained from nearby tile positions. The depth profiles corresponded to the expected deposition and erosion regions. The depth of the B-containing layer varied from tens of nanometers to several micrometers. The ablation rates of the deposit layers were generally 50–100 nm per laser shot, comparable to the rates of bulk W and Mo layers. The rate was considerably higher in the thickest deposits, which had stratified structures. The study shows that the LIBS method is sufficiently sensitive and has adequate depth resolution to study the B composition in the deposited layers.
{"title":"Ex-situ LIBS study for the determination of boron content in WEST divertor tiles after the 2019 campaign","authors":"Indrek Jõgi , Peeter Paris , Marta Malin Muru , Mathilde Diez , Emmanuelle Tsitrone , Jari Likonen , Antti Hakola , Eduard Grigore , the WEST team","doi":"10.1016/j.nme.2025.102004","DOIUrl":"10.1016/j.nme.2025.102004","url":null,"abstract":"<div><div>Laser Induced Breakdown Spectroscopy (LIBS) is a chemical analysis method that has been used in various remote handling applications and is currently being developed for the remote composition analysis of layers deposited on the first walls and divertor components of fusion reactors. The technique applies laser pulses to ablate a small amount of material, which forms a plasma plume and emits a spectrum characterizing the elements originating from the investigated material. The use of consecutive laser pulses at the same spot allows us to obtain elemental depth profiles with a resolution depending on the laser ablation rate. The aim of the present study was to investigate the feasibility of the LIBS method to analyze the boron (B) content in the surface layer of WEST divertor marker tiles removed after the 2019 experimental campaign (C4) when they were subjected to 16 boronizations, 3 during the C3 campaign and 13 during the C4 campaign. Boronizations are used in tungsten wall tokamaks to reduce the W contamination in core plasma, but the formed boron layers retain their effect only for a limited time due to erosion and redeposition. The six studied samples originated from different poloidal locations along the divertor including erosion and deposition-dominated regions. The LIBS spectra contained the emission lines of tungsten (W), molybdenum (Mo) and carbon (C), which were the main elements of the marker tile layers, and B as one of the species in the surface layers deposited during the experimental campaigns. The depth profiles showing the LIBS line intensities as a function of the applied laser pulse number at the same spot were generally consistent with the Glow Discharge Optical Emission Spectroscopy and Secondary Ion Mass Spectrometry depth profiles obtained from nearby tile positions. The depth profiles corresponded to the expected deposition and erosion regions. The depth of the B-containing layer varied from tens of nanometers to several micrometers. The ablation rates of the deposit layers were generally 50–100 nm per laser shot, comparable to the rates of bulk W and Mo layers. The rate was considerably higher in the thickest deposits, which had stratified structures. The study shows that the LIBS method is sufficiently sensitive and has adequate depth resolution to study the B composition in the deposited layers.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102004"},"PeriodicalIF":2.7,"publicationDate":"2025-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321265","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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