Pub Date : 2025-12-31DOI: 10.1016/j.nme.2025.102057
D. Marzullo , A. Clagnan , V.G. Belardi , A. Cardella , V. Imbriani , G. Mazzone , J.H. You
In the context of EUROfusion activities for the development of the DEMO reactor design, the divertor configuration is a major challenge. The current conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. The evaluations developed to identify the best water coolant thermal–hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes) led to the identification of a new divertor baseline solution, based on the new cooling water operating conditions, named Divertor Single Null High-Temperature (SNHT). Such conditions require water at relatively high temperature (295 °C) and pressure (15.5 MPa), posing new challenging issues related to the general layout of the divertor cassette, its structural robustness and the manufacturing technologies.
This work presents a comparative assessment between two different solutions proposed for the design and manufacturing of the divertor cassette body. A preliminary structural assessment and technological parameters are considered, as well as shielding and thermo-hydraulic performances.
{"title":"Design progress of EU DEMO divertor cassette","authors":"D. Marzullo , A. Clagnan , V.G. Belardi , A. Cardella , V. Imbriani , G. Mazzone , J.H. You","doi":"10.1016/j.nme.2025.102057","DOIUrl":"10.1016/j.nme.2025.102057","url":null,"abstract":"<div><div>In the context of EUROfusion activities for the development of the DEMO reactor design, the divertor configuration is a major challenge. The current conceptual divertor design is based on the use of EUROFER97 for the divertor cassette body, while tungsten monoblocks bonded to CuCrZr pipes are used for plasma-facing targets. The evaluations developed to identify the best water coolant thermal–hydraulic conditions avoiding material embrittlement (for EUROFER 97) and softening/hardening (for copper alloy pipes) led to the identification of a new divertor baseline solution, based on the new cooling water operating conditions, named Divertor Single Null High-Temperature (SNHT). Such conditions require water at relatively high temperature (295 °C) and pressure (15.5 MPa), posing new challenging issues related to the general layout of the divertor cassette, its structural robustness and the manufacturing technologies.</div><div>This work presents a comparative assessment between two different solutions proposed for the design and manufacturing of the divertor cassette body. A preliminary structural assessment and technological parameters are considered, as well as shielding and thermo-hydraulic performances.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102057"},"PeriodicalIF":2.7,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926748","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 : 2025-12-29DOI: 10.1016/j.nme.2025.102055
Quan-Fu Han, Jinxin Chen, Aoyu Mo, Wenjie Li, Haijun Li, Xiaowei Ma, Yunshan Xiong, Peng Shao, Bo Li, Kun Jie Yang, Yue-Lin Liu
Based on comprehensive first-principles calculations, this study systematically investigates hydrogen (H) segregation behavior at Σ3(112)[110] and Σ5(310)[100] tungsten (W) grain boundaries (GBs) under uniaxial tensile strain, and its consequent impact on GB mechanical strengths. Our results indicate that the application of tensile strain significantly promotes H segregation to both pristine and vacancy-containing GBs. This behavior is mainly attributed to the reduction of the local charge density at interstitial sites, thereby revealing a possible positive correlation between H segregation energy and local charge density. Comparing the H segregation behavior at the two pristine GBs, H preferentially segregates to the Σ5(310)[100] GB due to its lower charge density. However, this tendency is influenced by the formation of vacancies in the GB, and as the number of H atoms in the vacancies increases, the segregation behavior shifts. First-principles tensile tests show that segregated H substantially reduces the ultimate tensile strength of W GBs. This embrittlement effect intensifies with increasing H concentration and is particularly pronounced when H atoms localize directly on the GB plane. Moreover, the presence of vacancy-H clusters further degrades mechanical strength, especially in the Σ5(310)[100] GB. These findings highlight the critical role of mechanical strain in accelerating H embrittlement in W, providing essential insights for designing radiation-resistant plasma-facing materials in fusion reactors.
{"title":"First-principles study on hydrogen segregation in tungsten grain boundaries and its impact on their mechanical strengths: Uniaxial tensile strain effect","authors":"Quan-Fu Han, Jinxin Chen, Aoyu Mo, Wenjie Li, Haijun Li, Xiaowei Ma, Yunshan Xiong, Peng Shao, Bo Li, Kun Jie Yang, Yue-Lin Liu","doi":"10.1016/j.nme.2025.102055","DOIUrl":"10.1016/j.nme.2025.102055","url":null,"abstract":"<div><div>Based on comprehensive first-principles calculations, this study systematically investigates hydrogen (H) segregation behavior at Σ3(112)[110] and Σ5(310)[100] tungsten (W) grain boundaries (GBs) under uniaxial tensile strain, and its consequent impact on GB mechanical strengths. Our results indicate that the application of tensile strain significantly promotes H segregation to both pristine and vacancy-containing GBs. This behavior is mainly attributed to the reduction of the local charge density at interstitial sites, thereby revealing a possible positive correlation between H segregation energy and local charge density. Comparing the H segregation behavior at the two pristine GBs, H preferentially segregates to the Σ5(310)[100] GB due to its lower charge density. However, this tendency is influenced by the formation of vacancies in the GB, and as the number of H atoms in the vacancies increases, the segregation behavior shifts. First-principles tensile tests show that segregated H substantially reduces the ultimate tensile strength of W GBs. This embrittlement effect intensifies with increasing H concentration and is particularly pronounced when H atoms localize directly on the GB plane. Moreover, the presence of vacancy-H clusters further degrades mechanical strength, especially in the Σ5(310)[100] GB. These findings highlight the critical role of mechanical strain in accelerating H embrittlement in W, providing essential insights for designing radiation-resistant plasma-facing materials in fusion reactors.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102055"},"PeriodicalIF":2.7,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926751","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 : 2025-12-22DOI: 10.1016/j.nme.2025.102052
Jing Liang , Yu Li , Chen-Yuan Zhang , Si-Xin Lv , Chang Xu , Long-Qiang Han , Yi-Wen Zhu , Zhong-Shi Yang , Fang Ding , Guang-Nan Luo , Hai-Shan Zhou
The erosion of the tungsten (W) first wall by the seeding impurity neon (Ne) is foreseen in ITER. Accurate physical sputtering yields are crucial in defining the operating window that is consistent with the operational budget of the ITER divertor/main wall. However, the influence of crystal orientation and surface nanostructure—due to helium plasma exposure, on the physical sputtering yield is poorly understood. Here, we explore such influence for W bombarded by fusion-relevant Ne plasmas experimentally. In the first set of experiments, polished polycrystalline W targets were exposed to ∼ 50 eV Ne plasmas to a fluence of ∼ 3×1026 m−2. Subsequent secondary electron imaging revealed pronounced selective surface erosion. Combined with electron backscatter diffraction, we found that the (111) grains were more resilient to physical sputtering than the (100) grains. In the second set of experiments, He plasma exposure was performed to generate ‘fuzzy’ surfaces prior to Ne plasma exposure. By monitoring the intensity ratio between the W I and Ne II emission lines, strongly reduced, nonlinear erosion of the ‘fuzzy’ surfaces was observed. Measurable physical sputtering yields as low as 20 % of the smooth counterpart were recorded, which decreased with increasing ‘fuzzy’ layer thickness. The results highlight the impact of grain orientation and surface nanostructure on the physical sputtering yield of W bombarded by Ne. Moreover, the sputtering resistance of the ‘fuzzy’ layer may be exploited to boost the first wall performance in fusion devices.
{"title":"Grain orientation and surface nanostructure impact physical sputtering of tungsten by neon plasmas","authors":"Jing Liang , Yu Li , Chen-Yuan Zhang , Si-Xin Lv , Chang Xu , Long-Qiang Han , Yi-Wen Zhu , Zhong-Shi Yang , Fang Ding , Guang-Nan Luo , Hai-Shan Zhou","doi":"10.1016/j.nme.2025.102052","DOIUrl":"10.1016/j.nme.2025.102052","url":null,"abstract":"<div><div>The erosion of the tungsten (W) first wall by the seeding impurity neon (Ne) is foreseen in ITER. Accurate physical sputtering yields are crucial in defining the operating window that is consistent with the operational budget of the ITER divertor/main wall. However, the influence of crystal orientation and surface nanostructure—due to helium plasma exposure, on the physical sputtering yield is poorly understood. Here, we explore such influence for W bombarded by fusion-relevant Ne plasmas experimentally. In the first set of experiments, polished polycrystalline W targets were exposed to ∼ 50 eV Ne plasmas to a fluence of ∼ 3×10<sup>26</sup> m<sup>−2</sup>. Subsequent secondary electron imaging revealed pronounced selective surface erosion. Combined with electron backscatter diffraction, we found that the (111) grains were more resilient to physical sputtering than the (100) grains. In the second set of experiments, He plasma exposure was performed to generate ‘fuzzy’ surfaces prior to Ne plasma exposure. By monitoring the intensity ratio between the W I and Ne II emission lines, strongly reduced, nonlinear erosion of the ‘fuzzy’ surfaces was observed. Measurable physical sputtering yields as low as 20 % of the smooth counterpart were recorded, which decreased with increasing ‘fuzzy’ layer thickness. The results highlight the impact of grain orientation and surface nanostructure on the physical sputtering yield of W bombarded by Ne. Moreover, the sputtering resistance of the ‘fuzzy’ layer may be exploited to boost the first wall performance in fusion devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102052"},"PeriodicalIF":2.7,"publicationDate":"2025-12-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926682","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 : 2025-12-22DOI: 10.1016/j.nme.2025.102053
D.N. Gautam , D. Primetzhofer , M. Rubel , E. Pitthan
The retention of two neon isotopes, 20Ne and 22Ne, was studied by ion beam analysis (IBA) for thin-films of mixed W and B as well as for pure W and B layers grown on silicon-and tungsten-substrates by means of magnetron sputter deposition. Each isotope was implanted to a fluence of 3 × 1016 at./cm2 but at different energies (35–190 keV) to obtain deposition profiles closer to the surface and deeper into the film, depending on isotope and thin-film composition. Thermal annealing in combination with IBA was used to investigate the Ne-retention in a range of temperatures between RT and 1000 °C. Time-of-flight elastic recoil detection analysis was employed to monitor the retention and depth profiles of the Ne isotopes. Both Ne-isotopes remain at their original implantation depth, thus not indicating diffusion, intermixing or desorption for the full range of temperatures and for all studied compositions.
{"title":"Neon retention in tungsten, boron and mixed thin-films under the effects of thermal annealing studied by isotopic tracing","authors":"D.N. Gautam , D. Primetzhofer , M. Rubel , E. Pitthan","doi":"10.1016/j.nme.2025.102053","DOIUrl":"10.1016/j.nme.2025.102053","url":null,"abstract":"<div><div>The retention of two neon isotopes, <sup>20</sup>Ne and <sup>22</sup>Ne, was studied by ion beam analysis (IBA) for thin-films of mixed W and B as well as for pure W and B layers grown on silicon-and tungsten-substrates by means of magnetron sputter deposition. Each isotope was implanted to a fluence of 3 × 10<sup>16</sup> at./cm<sup>2</sup> but at different energies (35–190 keV) to obtain deposition profiles closer to the surface and deeper into the film, depending on isotope and thin-film composition. Thermal annealing in combination with IBA was used to investigate the Ne-retention in a range of temperatures between RT and 1000 °C. Time-of-flight elastic recoil detection analysis was employed to monitor the retention and depth profiles of the Ne isotopes. Both Ne-isotopes remain at their original implantation depth, thus not indicating diffusion, intermixing or desorption for the full range of temperatures and for all studied compositions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102053"},"PeriodicalIF":2.7,"publicationDate":"2025-12-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926683","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 : 2025-12-14DOI: 10.1016/j.nme.2025.102049
A.S. Teimane , E. Pajuste , L. Avotina , A. Lescinskis , A. Vitins , A.E. Goldmane , M. Sondars , R.J. Zabolockis , J. Likonen , A. Widdowson , JET Contributors
Tritium retention is a critical aspect of plasma-facing wall component performance in fusion reactors as well as reactor safety due to radiological risks it may pose. It is also of importance in the case of tungsten, including tungsten composites, which are selected as first wall and divertor material at devices such as ITER due to its high melting point and mechanical strength. This study aims to investigate surface characteristics, tritium retention behaviour and effect of baking on tungsten composite plasma-facing wall components from Joint European Torus (JET) divertor region and contribute to the understanding of tritium trapping within them.
Three ITER-like wall (ILW) experimental campaigns involved exposing tungsten-molybdenum coated carbon fibre composite (CFC) samples to deuterium-deuterium (D-D) plasma discharges at various operating conditions, including different plasma densities, temperatures, and exposure times. The plasma-facing surfaces were characterized using scanning electron microscopy (SEM) in combination with energy-dispersive x-ray spectroscopy (EDX) and tritium retention was assessed using thermal desorption spectroscopy (TDS) and full combustion. Baking cycle was simulated by keeping the sample at 350℃ for 100 h, followed by TDS and full combustion.
Results indicate tritium retention varying from 2 to 120∙1012 T atoms/plasma facing surface cm2. A deposition layer was found to be present for most samples analysed in this study ranging from 0 to 58 µm in thickness. For Tile 0 an increase in tritium retention was observed by the increase in the thickness of the deposition layer, whilst for Tile 1 deposition was not found to be the main source of retention. Tritium desorption temperatures were found to be higher than that proposed for baking at ITER − for Tile 0 tritium desorption peaks at about 540-640℃, while for tile 1 it is generally lower, but with a larger deviation ranging from 350 up to 570℃.
{"title":"Investigating tritium retention in tungsten coated plasma facing components from the divertor region of the Joint European Torus (JET) after ITER like-wall campaigns","authors":"A.S. Teimane , E. Pajuste , L. Avotina , A. Lescinskis , A. Vitins , A.E. Goldmane , M. Sondars , R.J. Zabolockis , J. Likonen , A. Widdowson , JET Contributors","doi":"10.1016/j.nme.2025.102049","DOIUrl":"10.1016/j.nme.2025.102049","url":null,"abstract":"<div><div>Tritium retention is a critical aspect of plasma-facing wall component performance in fusion reactors as well as reactor safety due to radiological risks it may pose. It is also of importance in the case of tungsten, including tungsten composites, which are selected as first wall and divertor material at devices such as ITER due to its high melting point and mechanical strength. This study aims to investigate surface characteristics, tritium retention behaviour and effect of baking on tungsten composite plasma-facing wall components from Joint European Torus (JET) divertor region and contribute to the understanding of tritium trapping within them.</div><div>Three ITER-like wall (ILW) experimental campaigns involved exposing tungsten-molybdenum coated carbon fibre composite (CFC) samples to deuterium-deuterium (D-D) plasma discharges at various operating conditions, including different plasma densities, temperatures, and exposure times. The plasma-facing surfaces were characterized using scanning electron microscopy (SEM) in combination with energy-dispersive x-ray spectroscopy (EDX) and tritium retention was assessed using thermal desorption spectroscopy (TDS) and full combustion. Baking cycle was simulated by keeping the sample at 350℃ for 100 h, followed by TDS and full combustion.</div><div>Results indicate tritium retention varying from 2 to 120∙10<sup>12</sup> T atoms/plasma facing surface cm<sup>2</sup>. A deposition layer was found to be present for most samples analysed in this study ranging from 0 to 58 µm in thickness. For Tile 0 an increase in tritium retention was observed by the increase in the thickness of the deposition layer, whilst for Tile 1 deposition was not found to be the main source of retention. Tritium desorption temperatures were found to be higher than that proposed for baking at ITER − for Tile 0 tritium desorption peaks at about 540-640℃, while for tile 1 it is generally lower, but with a larger deviation ranging from 350 up to 570℃.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102049"},"PeriodicalIF":2.7,"publicationDate":"2025-12-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791723","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 : 2025-12-14DOI: 10.1016/j.nme.2025.102050
R. Mateus , N. Catarino , E. Alves , M. Diez , E. Bernard , E. Tsitrone , M. Balden , M. Mayer , J. Likonen , A. Hakola , the WEST team
Erosion marker tiles mounted in the lower divertor of WEST were exposed during Phase 1 of plasma operations to evaluate poloidal erosion and re-deposition profiles on the tiles. Previous analyses performed to the exposed tiles have shown distinct erosion- or deposition-dominated patterns on them. Afterwards, core-drilled disks cut from the tiles were sent to different laboratories for further and detailed analysis. The present work relates the main results achieved from five characteristic regions of the tiles after completion of the C3, C4, and C5 experimental campaigns on WEST. SIMS and complementary IBA measurements were carried out and the corresponding elemental depth profiles strongly agree, confirming the main earlier conclusions. Deposits are composed of 2H, B, C, O, Mo and W, mainly. Low amounts of Cr, Fe, Ni and Cu were identified as additional metallic impurities. The research confirmed the locations of thin deposition zones nearby the inner and outer divertor limits: at the inner region, the deposition of B and C is particularly enhanced after C4 and C5. Strong erosion zones are located at the inner and outer strike point (ISP and OSP, respectively) areas: only a small erosion occurred after C3, which evolved after C4; nevertheless, the deposition of B and C is enhanced at the OSP edge after C5 nearby the thin deposition zone. Thick deposits appear in the neighborhood of ISP, towards the high field side, and evolve significantly after C4. The amount of O follows the deposition of B. Low retained amounts of 2H were quantified.
{"title":"Elemental analysis of divertor marker tiles exposed during the 2018 (C3), 2019 (C4) and 2020 (C5) WEST campaigns","authors":"R. Mateus , N. Catarino , E. Alves , M. Diez , E. Bernard , E. Tsitrone , M. Balden , M. Mayer , J. Likonen , A. Hakola , the WEST team","doi":"10.1016/j.nme.2025.102050","DOIUrl":"10.1016/j.nme.2025.102050","url":null,"abstract":"<div><div>Erosion marker tiles mounted in the lower divertor of WEST were exposed during Phase 1 of plasma operations to evaluate poloidal erosion and re-deposition profiles on the tiles. Previous analyses performed to the exposed tiles have shown distinct erosion- or deposition-dominated patterns on them. Afterwards, core-drilled disks cut from the tiles were sent to different laboratories for further and detailed analysis. The present work relates the main results achieved from five characteristic regions of the tiles after completion of the C3, C4, and C5 experimental campaigns on WEST. SIMS and complementary IBA measurements were carried out and the corresponding elemental depth profiles strongly agree, confirming the main earlier conclusions. Deposits are composed of <sup>2</sup>H, B, C, O, Mo and W, mainly. Low amounts of Cr, Fe, Ni and Cu were identified as additional metallic impurities. The research confirmed the locations of thin deposition zones nearby the inner and outer divertor limits: at the inner region, the deposition of B and C is particularly enhanced after C4 and C5. Strong erosion zones are located at the inner and outer strike point (ISP and OSP, respectively) areas: only a small erosion occurred after C3, which evolved after C4; nevertheless, the deposition of B and C is enhanced at the OSP edge after C5 nearby the thin deposition zone. Thick deposits appear in the neighborhood of ISP, towards the high field side, and evolve significantly after C4. The amount of O follows the deposition of B. Low retained amounts of <sup>2</sup>H were quantified.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102050"},"PeriodicalIF":2.7,"publicationDate":"2025-12-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791720","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 : 2025-12-13DOI: 10.1016/j.nme.2025.102048
L. Vignitchouk, JET Contributors
Multiphase Navier–Stokes simulations of castellated beryllium plates exposed to JET-like disruption plasma loads are performed to investigate melt transport in the vicinity of gaps and the formation of re-solidified bridges between adjacent castellation blocks. It is found that two-dimensional computations are able to predict whether bridging occurs and that they agree with experimental data in terms of characteristic melt infiltration depths and global material transport along the surface. However, three-dimensional set-ups appear to be necessary when estimates of the damaged component’s surface morphology are sought in cases where bridging does not occur. Comparisons with simplified shallow-water models confirm that such models are applicable to scenarios in which bridges have already been formed, although they tend to overestimate the net melt displacement.
{"title":"Simulations of beryllium castellation gap bridging during vertical displacement events","authors":"L. Vignitchouk, JET Contributors","doi":"10.1016/j.nme.2025.102048","DOIUrl":"10.1016/j.nme.2025.102048","url":null,"abstract":"<div><div>Multiphase Navier–Stokes simulations of castellated beryllium plates exposed to JET-like disruption plasma loads are performed to investigate melt transport in the vicinity of gaps and the formation of re-solidified bridges between adjacent castellation blocks. It is found that two-dimensional computations are able to predict whether bridging occurs and that they agree with experimental data in terms of characteristic melt infiltration depths and global material transport along the surface. However, three-dimensional set-ups appear to be necessary when estimates of the damaged component’s surface morphology are sought in cases where bridging does not occur. Comparisons with simplified shallow-water models confirm that such models are applicable to scenarios in which bridges have already been formed, although they tend to overestimate the net melt displacement.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102048"},"PeriodicalIF":2.7,"publicationDate":"2025-12-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791719","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 : 2025-12-12DOI: 10.1016/j.nme.2025.102047
Al-Montaser Bellah A. Al-Ajlony , Ghadeer H. Al-Malkawi
This study presents a comprehensive Monte Carlo simulation-based investigation into the angle of incidence dependence of physical and chemical sputtering of graphite targets irradiated by He+, T+, and D+ ions in the near-threshold energy regime. This investigation has been executed by utilizing two advanced simulation codes, RDS-BASIC and SDTrimSP. In this study, we assess sputtering yields and energy thresholds across incidence angles ranging from 0° to 80°. Results indicate that physical sputtering yields are maximized at grazing angles (60°–80°), accompanied by a systematic decrease in sputtering threshold energy with increasing angle. In contrast, chemical sputtering, modeled for T+ and D+ ions, reveals a two-step threshold behavior: an initial erosion onset at ∼ 5 eV, and a secondary enhancement between 8–13 eV driven by physical displacement effects. Which also was found to influence the angle og incidence dependence of the chemical sputtering. Thereby, to physical sputtering, chemical erosion exhibits a peak at intermediate angles (60°–70°). These findings offer key insights into the erosion mechanisms of plasma-facing components and support the optimized design of carbon-based materials for future nuclear fusion reactor applications.
{"title":"Effect of the angle of incidence of He, T, and D ions irradiation on physical and chemical sputtering of graphite targets in the near sputtering threshold energy regime","authors":"Al-Montaser Bellah A. Al-Ajlony , Ghadeer H. Al-Malkawi","doi":"10.1016/j.nme.2025.102047","DOIUrl":"10.1016/j.nme.2025.102047","url":null,"abstract":"<div><div>This study presents a comprehensive Monte Carlo simulation-based investigation into the angle of incidence dependence of physical and chemical sputtering of graphite targets irradiated by He<sup>+</sup>, T<sup>+</sup>, and D<sup>+</sup> ions in the near-threshold energy regime. This investigation has been executed by utilizing two advanced simulation codes, RDS-BASIC and SDTrimSP. In this study, we assess sputtering yields and energy thresholds across incidence angles ranging from 0° to 80°. Results indicate that physical sputtering yields are maximized at grazing angles (60°–80°), accompanied by a systematic decrease in sputtering threshold energy with increasing angle. In contrast, chemical sputtering, modeled for T<sup>+</sup> and D<sup>+</sup> ions, reveals a two-step threshold behavior: an initial erosion onset at ∼ 5 eV, and a secondary enhancement between 8–13 eV driven by physical displacement effects. Which also was found to influence the angle og incidence dependence of the chemical sputtering. Thereby, to physical sputtering, chemical erosion exhibits a peak at intermediate angles (60°–70°). These findings offer key insights into the erosion mechanisms of plasma-facing components and support the optimized design of carbon-based materials for future nuclear fusion reactor applications.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102047"},"PeriodicalIF":2.7,"publicationDate":"2025-12-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791721","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 : 2025-12-11DOI: 10.1016/j.nme.2025.102043
Yin Zhong Shen, Sung Ho Kim, Sung Hwan Yeo
The precipitate phases of an 11Cr F/M steel in normalized-and-tempered state and after irradiation with 3.5 MeV Fe13+ ions at 700 °C to 0.84 dpa were studied using transmission electron microscopy. Cr-rich precipitate phase underwent great changes under irradiation. While the existing Cr-rich M23C6 precipitates with a Cr/Fe ratio of about 1.8 remained, Cr-rich M23C6 precipitates with a Cr/Fe ratio of about 1.0 were precipitated during irradiation. Irradiation induced the precipitation of rod-like Cr-rich M23C6 precipitates within martensite laths, and Cr-rich M23C6 precipitates with a parallelogram morphology in the matrix. Irradiation-induced Cr-rich M2X (Cr2N type) carbonitride and Cr-rich M7C3 carbide were identified. Irradiation significantly affected δ-ferrite in the steel, resulting in the precipitation of dense and small Cr-rich M3X2 carbonitrides based on Cr3C2 within δ-ferrite grains. Irradiation caused significant changes in the nature of Nb-rich precipitate phases. While the existing Nb-Ta-rich MX carbonitrides based on NbC (fcc lattice, a = 0.4469 nm) remained under irradiation, irradiation induced three types of Nb-rich phases, including Nb-Ta-rich MX carbonitride based on NbC (fcc lattice, a = 1.115 nm), Nb2C (simple orthorhombic lattice) carbide, and Nb-Ta-rich M6X5 carbonitride based on Nb6C5 (base-centered monoclinic lattice). Irradiation also induced the formation of two types of intermetallic compound phases, σ-FeCrW (base-centered tetragonal lattice) and Fe-Cr (bcc lattice) which are absent in the normalized-and-tempered steel, within the δ-ferrite adjacent to δ-ferrite-martensite boundaries. The formation of the irradiation-induced precipitate phases is also discussed.
{"title":"Fe-ion irradiation induced changes in precipitates of 11Cr ferritic/martensitic steel","authors":"Yin Zhong Shen, Sung Ho Kim, Sung Hwan Yeo","doi":"10.1016/j.nme.2025.102043","DOIUrl":"10.1016/j.nme.2025.102043","url":null,"abstract":"<div><div>The precipitate phases of an 11Cr F/M steel in normalized-and-tempered state and after irradiation with 3.5 MeV Fe<sup>13+</sup> ions at 700 °C to 0.84 dpa were studied using transmission electron microscopy. Cr-rich precipitate phase underwent great changes under irradiation. While the existing Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a Cr/Fe ratio of about 1.8 remained, Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a Cr/Fe ratio of about 1.0 were precipitated during irradiation. Irradiation induced the precipitation of rod-like Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates within martensite laths, and Cr-rich M<sub>23</sub>C<sub>6</sub> precipitates with a parallelogram morphology in the matrix. Irradiation-induced Cr-rich M<sub>2</sub>X (Cr<sub>2</sub>N type) carbonitride and Cr-rich M<sub>7</sub>C<sub>3</sub> carbide were identified. Irradiation significantly affected δ-ferrite in the steel, resulting in the precipitation of dense and small Cr-rich M<sub>3</sub>X<sub>2</sub> carbonitrides based on Cr<sub>3</sub>C<sub>2</sub> within δ-ferrite grains. Irradiation caused significant changes in the nature of Nb-rich precipitate phases. While the existing Nb-Ta-rich MX carbonitrides based on NbC (fcc lattice, a = 0.4469 nm) remained under irradiation, irradiation induced three types of Nb-rich phases, including Nb-Ta-rich MX carbonitride based on NbC (fcc lattice, a = 1.115 nm), Nb<sub>2</sub>C (simple orthorhombic lattice) carbide, and Nb-Ta-rich M<sub>6</sub>X<sub>5</sub> carbonitride based on Nb<sub>6</sub>C<sub>5</sub> (base-centered monoclinic lattice). Irradiation also induced the formation of two types of intermetallic compound phases, σ-FeCrW (base-centered tetragonal lattice) and Fe-Cr (bcc lattice) which are absent in the normalized-and-tempered steel, within the δ-ferrite adjacent to δ-ferrite-martensite boundaries. The formation of the irradiation-induced precipitate phases is also discussed.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102043"},"PeriodicalIF":2.7,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145926681","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 : 2025-12-11DOI: 10.1016/j.nme.2025.102042
A. Gallo , P. Manas , T. Wauters , M. Diez , E. Geulin , E. Hodille , J. Gaspar , N. Rivals , P. Puglia , Ph. Moreau , D. Douai , T. Alarcon , V. Anzallo , E. Caprin , M. De Combarieu , F.P. Pellissier , P. Devynck , C. Guillemaut , C. Desgranges , B. Guillermin , A. Hakola
The recent ITER re-baseline with the adoption of a full-W wall calls for mandatory boronization studies. ITER pulses will be inboard limited on the W tiles of the central column for several seconds during the current ramp up phase. Our first question of this study is: will it be possible to efficiently start plasma operations in a full-W ITER without any boronization? In particular, throughout the start of research operations (SRO), ITER will be equipped with an asymmetric boronization system as glow anodes in the equatorial plane will not be uniformly distributed in the toroidal direction due to the limited availability of ports. According to recent simulations, such arrangement of the glow anodes could lead to a strongly non-uniform B layer with depleted regions. Our second question hence is: should a boronization be needed to start plasma operations in ITER, would a non-uniform B layer be enough? In November 2024, we attempted to restart WEST plasma operations without boronization after a vent and after installing new bulk W limiter tiles. In about 4 days of operation corresponding to 74 pulse attempts, we reached a maximum pulse duration of 1.55 s and a maximum plasma current of 600 kA. Plasmas were cold and dense, mostly detached from the inboard limiter and dominated by light impurities with radiated power fractions close to unity. No runaway electron beams were observed but the restart without boronization was not timely. We then carried out the first WEST boronization utilizing only 3 out of 6 diborane (B2D6) inlets (half torus), to deposit a non-uniform B layer. Repeatable, 10 s long, ohmic limiter pulses were immediately achieved with radiated power fractions between 50 % and 70 %. Through a separate experiment in February 2025, we achieved matching pulses before and after a second non-uniform boronization to better characterize its effects: the radiated fraction initially dropped by 22 % with the reduction mainly occurring in the central plasma and well correlating with lower UV signals for O, N and W. These effects almost vanished by the end of the first day after the non-uniform boronization corresponding to a cumulated injected energy of 0.7 GJ.
{"title":"Effect of spatially non-uniform boronization on plasma restart in WEST","authors":"A. Gallo , P. Manas , T. Wauters , M. Diez , E. Geulin , E. Hodille , J. Gaspar , N. Rivals , P. Puglia , Ph. Moreau , D. Douai , T. Alarcon , V. Anzallo , E. Caprin , M. De Combarieu , F.P. Pellissier , P. Devynck , C. Guillemaut , C. Desgranges , B. Guillermin , A. Hakola","doi":"10.1016/j.nme.2025.102042","DOIUrl":"10.1016/j.nme.2025.102042","url":null,"abstract":"<div><div>The recent ITER re-baseline with the adoption of a full-W wall calls for mandatory boronization studies. ITER pulses will be inboard limited on the W tiles of the central column for several seconds during the current ramp up phase. Our first question of this study is: <em>will it be possible to efficiently start plasma operations in a full-W ITER without any boronization?</em> In particular, throughout the start of research operations (SRO), ITER will be equipped with an asymmetric boronization system as glow anodes in the equatorial plane will not be uniformly distributed in the toroidal direction due to the limited availability of ports. According to recent simulations, such arrangement of the glow anodes could lead to a strongly non-uniform B layer with depleted regions. Our second question hence is: <em>should a boronization be needed to start plasma operations in ITER, would a non-uniform B layer be enough?</em> In November 2024, we attempted to restart WEST plasma operations without boronization after a vent and after installing new bulk W limiter tiles. In about 4 days of operation corresponding to 74 pulse attempts, we reached a maximum pulse duration of 1.55 s and a maximum plasma current of 600 kA. Plasmas were cold and dense, mostly detached from the inboard limiter and dominated by light impurities with radiated power fractions close to unity. No runaway electron beams were observed but the restart without boronization was not timely. We then carried out the first WEST boronization utilizing only 3 out of 6 diborane (B<sub>2</sub>D<sub>6</sub>) inlets (half torus), to deposit a non-uniform B layer. Repeatable, 10 s long, ohmic limiter pulses were immediately achieved with radiated power fractions between 50 % and 70 %. Through a separate experiment in February 2025, we achieved matching pulses before and after a second non-uniform boronization to better characterize its effects: the radiated fraction initially dropped by 22 % with the reduction mainly occurring in the central plasma and well correlating with lower UV signals for O, N and W. These effects almost vanished by the end of the first day after the non-uniform boronization corresponding to a cumulated injected energy of 0.7 GJ.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102042"},"PeriodicalIF":2.7,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145791722","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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