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}
Pub Date : 2025-12-11DOI: 10.1016/j.nme.2025.102045
P. Lorusso , S. Roccella , M. Di Bartolomeo , E. Cacciotti , M. Cerocchi , R. De Luca , E. Martelli , L. Verdini , K. Hunger , P. Junghanns , A. von Müller , B. Böswirth , H. Greuner , J. Riesch , J.H. You
Among the R&D tasks undertaken for the technological development of plasma facing components, a research activity has been undertaken under the framework of the EUROfusion consortium to pursue the consolidation and verification of the current target concepts envisioned for DEMO, i.e., the ITER-like baseline concept and the back-up concept with tungsten-fiber-reinforced copper (Wf-Cu) pipes. Focus has been addressed on the back-up solution (Wf-Cu pipes) finding alternative technological solutions for monoblock-pipe joining in order to reduce the use of materials having high activation and/or degradation under neutron irradiation. Among the brazing alloys tested for the monoblock/Wf-Cu pipe joint, the Gemco commercial alloy has been selected as the most suitable, thanks to its good joining capability and low content of nickel, which suffers high neutron activation. However, the use of Wf-Cu pipes instead of standard CuCrZr ones complicates the joining. One reason could be that the lower thermal expansion coefficient of W compared to Cu affects the overall expansion behaviour of the tube, making it difficult to fill the gap between the surfaces to be joined (necessary for the assembly) during the brazing process. The present paper describes the manufacturing activities which led to the fabrication of small mock-ups. In particular, fabrication, testing at High Heat Flux (HHF) conditions and analyses of a small mock-up are presented. Non-destructive examinations by ultrasonic testing have been performed pre- and post-HHF tests to assess the structural integrity of the mock-up and the reliability of the joining. Furthermore, microscopic examinations at different magnifications have been carried out to highlight microstructural modifications, recrystallization, and potential defect propagation due to thermal cycling.
{"title":"Manufacturing and high heat flux testing of advanced target mock-ups for the EU-DEMO divertor target","authors":"P. Lorusso , S. Roccella , M. Di Bartolomeo , E. Cacciotti , M. Cerocchi , R. De Luca , E. Martelli , L. Verdini , K. Hunger , P. Junghanns , A. von Müller , B. Böswirth , H. Greuner , J. Riesch , J.H. You","doi":"10.1016/j.nme.2025.102045","DOIUrl":"10.1016/j.nme.2025.102045","url":null,"abstract":"<div><div>Among the R&D tasks undertaken for the technological development of plasma facing components, a research activity has been undertaken under the framework of the EUROfusion consortium to pursue the consolidation and verification of the current target concepts envisioned for DEMO, i.e., the ITER-like baseline concept and the back-up concept with tungsten-fiber-reinforced copper (Wf-Cu) pipes. Focus has been addressed on the back-up solution (Wf-Cu pipes) finding alternative technological solutions for monoblock-pipe joining in order to reduce the use of materials having high activation and/or degradation under neutron irradiation. Among the brazing alloys tested for the monoblock/Wf-Cu pipe joint, the Gemco commercial alloy has been selected as the most suitable, thanks to its good joining capability and low content of nickel, which suffers high neutron activation. However, the use of Wf-Cu pipes instead of standard CuCrZr ones complicates the joining. One reason could be that the lower thermal expansion coefficient of W compared to Cu affects the overall expansion behaviour of the tube, making it difficult to fill the gap between the surfaces to be joined (necessary for the assembly) during the brazing process. The present paper describes the manufacturing activities which led to the fabrication of small mock-ups. In particular, fabrication, testing at High Heat Flux (HHF) conditions and analyses of a small mock-up are presented. Non-destructive examinations by ultrasonic testing have been performed pre- and post-HHF tests to assess the structural integrity of the mock-up and the reliability of the joining. Furthermore, microscopic examinations at different magnifications have been carried out to highlight microstructural modifications, recrystallization, and potential defect propagation due to thermal cycling.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102045"},"PeriodicalIF":2.7,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145738858","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.2025.102044
Qilong Cao , Fangqing Qian , Xiaolin Li , Panpan Wang , He Tong , Yange Zhang , Yichun Xu , Xianping Wang , C.S. Liu , Xiang-Yan Li
Grain boundaries (GBs) serve as critical microstructural features in determining the radiation resistance of nanocrystalline metals through their capacity to absorb radiation-induced defects, particularly self-interstitial atoms (SIAs) and vacancies (Vs). Nevertheless, the fundamental mechanism governing SIA dynamics near GBs that dominate defect–GB interactions remain inadequately elucidated. This knowledge gap is primarily due to the fact that the time scale of SIA diffusion and its segregation towards GBs is significantly smaller than that of other events, leading to its preferential segregation to GBs and decoupling from other defects. In this study, we employ combined molecular statics and dynamics simulations to systematically investigate the atomic-scale behavior of SIA and its cluster (SIAn) states near GBs in three body-centered cubic metals (tungsten, molybdenum and iron), while concurrently exploring potential V annihilation mechanisms associated with these processes. We propose a “self-blocking” mechanism governing the behavior of the SIAn near GBs at low temperatures. As accumulated SIAns progressively occupy GB trapping sites, subsequent incoming clusters experience spatial confinement within the near-boundary region. These constrained SIAns exhibit intermediate energy states between bulk configuration states and fully GB-trapped states. Crucially, the partially constrained SIAns demonstrate two distinctive features of an enhanced V annihilation volume exceeding that of GB-trapped clusters by several times (attributed to their preserved bulk-like atomic configurations) and long-range repulsive interactions with bulk SIAs (potentially modifying defect-trapping dynamics of GBs). This self-blocking phenomenon implies both a saturation threshold for GB defect absorption capacity and an extended boundary zone facilitating V–SIA recombination. The spatially confined annihilation pathway near GBs provides a potential channel for eliminating bulk defects at low temperatures, offering new insights into the atomic-scale self-healing processes in nanocrystalline materials.
{"title":"Self-blocking driven interstitial confinement at metallic grain boundaries","authors":"Qilong Cao , Fangqing Qian , Xiaolin Li , Panpan Wang , He Tong , Yange Zhang , Yichun Xu , Xianping Wang , C.S. Liu , Xiang-Yan Li","doi":"10.1016/j.nme.2025.102044","DOIUrl":"10.1016/j.nme.2025.102044","url":null,"abstract":"<div><div>Grain boundaries (GBs) serve as critical microstructural features in determining the radiation resistance of nanocrystalline metals through their capacity to absorb radiation-induced defects, particularly self-interstitial atoms (SIAs) and vacancies (Vs). Nevertheless, the fundamental mechanism governing SIA dynamics near GBs that dominate defect–GB interactions remain inadequately elucidated. This knowledge gap is primarily due to the fact that the time scale of SIA diffusion and its segregation towards GBs is significantly smaller than that of other events, leading to its preferential segregation to GBs and decoupling from other defects. In this study, we employ combined molecular statics and dynamics simulations to systematically investigate the atomic-scale behavior of SIA and its cluster (SIA<sub>n</sub>) states near GBs in three body-centered cubic metals (tungsten, molybdenum and iron), while concurrently exploring potential V annihilation mechanisms associated with these processes. We propose a “self-blocking” mechanism governing the behavior of the SIA<sub>n</sub> near GBs at low temperatures. As accumulated SIA<sub>n</sub>s progressively occupy GB trapping sites, subsequent incoming clusters experience spatial confinement within the near-boundary region. These constrained SIA<sub>n</sub>s exhibit intermediate energy states between bulk configuration states and fully GB-trapped states. Crucially, the partially constrained SIA<sub>n</sub>s demonstrate two distinctive features of an enhanced V annihilation volume exceeding that of GB-trapped clusters by several times (attributed to their preserved bulk-like atomic configurations) and long-range repulsive interactions with bulk SIAs (potentially modifying defect-trapping dynamics of GBs). This self-blocking phenomenon implies both a saturation threshold for GB defect absorption capacity and an extended boundary zone facilitating V–SIA recombination. The spatially confined annihilation pathway near GBs provides a potential channel for eliminating bulk defects at low temperatures, offering new insights into the atomic-scale self-healing processes in nanocrystalline materials.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102044"},"PeriodicalIF":2.7,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145738857","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 : 2025-12-01DOI: 10.1016/j.nme.2025.102032
S. Saari , A. Hakola , J. Karhunen , M. Balden , C. Baumann , A. Järvinen , K. Krieger , H. Kumpulainen , J. Likonen , J. Romazanov , ASDEX Upgrade Team , EUROfusion Tokamak Exploitation Team
Erosion of small marker surfaces in an experiment conducted at the ASDEX Upgrade tokamak was modelled using the ERO2.0 code. In the experiment 5 × 5 mm2 and 1 × 1 mm2 Au marker spots were exposed to a series of high-temperature L-mode plasmas in the low-field side strike point region to serve as proxies for measuring net and gross erosion of W, respectively. An ERO2.0 simulation setup was created for the experiment using background plasma produced using OSM and new angle-dependent reflection and sputtering data for Au generated with the SDTrimSP code. The simulated net erosion of the Au markers agreed closely with the measured values. The erosion of the Au markers was induced mainly by the light B, C and N impurities defined as fixed concentrations in the background plasma. The Au markers were found to undergo up to 15–20 times stronger net erosion in comparison to a uniform W surface. This was attributed to 3–4 times stronger gross erosion of Au in comparison to W and deposition of the eroded Au mostly outside of the markers. Consequently, the simulations suggest strongly compromised capability of Au to act as proxy markers for W in erosion studies due to the significantly higher gross erosion yield of Au and insufficient size of the 5 × 5 mm2 markers for successful representation of net erosion, as eroded particles migrate along the plasma flow mostly outside the markers.
{"title":"ERO2.0 modelling of divertor marker erosion in ASDEX Upgrade L-mode experiments","authors":"S. Saari , A. Hakola , J. Karhunen , M. Balden , C. Baumann , A. Järvinen , K. Krieger , H. Kumpulainen , J. Likonen , J. Romazanov , ASDEX Upgrade Team , EUROfusion Tokamak Exploitation Team","doi":"10.1016/j.nme.2025.102032","DOIUrl":"10.1016/j.nme.2025.102032","url":null,"abstract":"<div><div>Erosion of small marker surfaces in an experiment conducted at the ASDEX Upgrade tokamak was modelled using the ERO2.0 code. In the experiment 5 × 5 mm<sup>2</sup> and 1 × 1 mm<sup>2</sup> Au marker spots were exposed to a series of high-temperature L-mode plasmas in the low-field side strike point region to serve as proxies for measuring net and gross erosion of W, respectively. An ERO2.0 simulation setup was created for the experiment using background plasma produced using OSM and new angle-dependent reflection and sputtering data for Au generated with the SDTrimSP code. The simulated net erosion of the Au markers agreed closely with the measured values. The erosion of the Au markers was induced mainly by the light B, C and N impurities defined as fixed concentrations in the background plasma. The Au markers were found to undergo up to 15–20 times stronger net erosion in comparison to a uniform W surface. This was attributed to 3–4 times stronger gross erosion of Au in comparison to W and deposition of the eroded Au mostly outside of the markers. Consequently, the simulations suggest strongly compromised capability of Au to act as proxy markers for W in erosion studies due to the significantly higher gross erosion yield of Au and insufficient size of the 5 × 5 mm<sup>2</sup> markers for successful representation of net erosion, as eroded particles migrate along the plasma flow mostly outside the markers.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102032"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145624005","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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