Pub Date : 2025-09-18DOI: 10.1016/j.nme.2025.101988
Rongrui Li , Guillermo Álvarez , Ayla Ipakchi , Livia Cupertino-Malheiros , Mark R. Gilbert , Emilio Martínez-Pañeda , Eric Prestat
The oxidation of pure W and the sublimation of W oxide have been investigated to assess their impact on the lifecycle of a fusion power plant. Pure W has been oxidised at temperatures between 400 and 1050 °C and for durations ranging between 1 and 70 h. The formation of voids and cracks has been observed at temperatures above 600 °C, leading to the formation of dust or oxide spalling, which could be problematic in maintenance and waste-handling scenarios of a fusion power plant. Preferential oxidation taking place at the edge of the specimen was characterised, and its impact is discussed in relation to component design. Characterisation using electron microscopy and Raman spectroscopy revealed that the oxide scale is formed of three main layers: the inner layer is 30–50 nm thick WO oxide, the middle layer is a 10–20 m thick of WO2.72 and the outer layer is formed of WO2.9/WO phases — whose thickness varies according to the total thickness of the oxide scale. The observed microstructure is discussed in relation to the parabolic-to-linear kinetics and its potential impact on tritium permeation and detritiation efficiency.
{"title":"Understanding the oxidation of pure tungsten in air and its impact on the lifecycle of a fusion power plant","authors":"Rongrui Li , Guillermo Álvarez , Ayla Ipakchi , Livia Cupertino-Malheiros , Mark R. Gilbert , Emilio Martínez-Pañeda , Eric Prestat","doi":"10.1016/j.nme.2025.101988","DOIUrl":"10.1016/j.nme.2025.101988","url":null,"abstract":"<div><div>The oxidation of pure W and the sublimation of W oxide have been investigated to assess their impact on the lifecycle of a fusion power plant. Pure W has been oxidised at temperatures between 400 and 1050 °C and for durations ranging between 1 and 70 h. The formation of voids and cracks has been observed at temperatures above 600 °C, leading to the formation of dust or oxide spalling, which could be problematic in maintenance and waste-handling scenarios of a fusion power plant. Preferential oxidation taking place at the edge of the specimen was characterised, and its impact is discussed in relation to component design. Characterisation using electron microscopy and Raman spectroscopy revealed that the oxide scale is formed of three main layers: the inner layer is 30–50 nm thick WO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> oxide, the middle layer is a 10–20 <span><math><mi>μ</mi></math></span>m thick of WO<sub>2.72</sub> and the outer layer is formed of WO<sub>2.9</sub>/WO<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span> phases — whose thickness varies according to the total thickness of the oxide scale. The observed microstructure is discussed in relation to the parabolic-to-linear kinetics and its potential impact on tritium permeation and detritiation efficiency.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101988"},"PeriodicalIF":2.7,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221713","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 : 2025-09-17DOI: 10.1016/j.nme.2025.101987
Ryan D. Kerr , Duc Nguyen-Manh , Mark R. Gilbert , Samuel T. Murphy
The addition of Cr and Y into tungsten can dramatically increase the oxidation resistance of the first wall of a future fusion reactor, thereby reducing the risk of formation of volatile WO and the release of radioactive material. Experimental observations suggest that in these SMART alloys, yttrium facilitates the formation of a self-passivating layer of Cr2O3 at the metal surface, however, how exactly the Y does this remains unclear. Therefore, this work explores the phase stability of compounds consisting of W–Y–Cr–O and solution energies for the different components in tungsten using density functional theory. The simulations suggest that there is a substantial thermodynamic driving force for the formation of YO, especially from yttrium and oxygen solvated in bulk tungsten. These observations suggest that the role of the yttrium may be to remove the oxygen that may inhibit Cr diffusion to the surface from the tungsten grains. This observation is in accordance with experimental studies showing that the oxidation resistance in the alloy occurs when the oxygen–yttrium ratio in the alloy is close to the stoichiometric ratio for YO.
{"title":"Stability of oxide phases in W–Cr–Y SMART alloys","authors":"Ryan D. Kerr , Duc Nguyen-Manh , Mark R. Gilbert , Samuel T. Murphy","doi":"10.1016/j.nme.2025.101987","DOIUrl":"10.1016/j.nme.2025.101987","url":null,"abstract":"<div><div>The addition of Cr and Y into tungsten can dramatically increase the oxidation resistance of the first wall of a future fusion reactor, thereby reducing the risk of formation of volatile WO<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span> and the release of radioactive material. Experimental observations suggest that in these SMART alloys, yttrium facilitates the formation of a self-passivating layer of Cr<sub>2</sub>O<sub>3</sub> at the metal surface, however, how exactly the Y does this remains unclear. Therefore, this work explores the phase stability of compounds consisting of W–Y–Cr–O and solution energies for the different components in tungsten using density functional theory. The simulations suggest that there is a substantial thermodynamic driving force for the formation of Y<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>, especially from yttrium and oxygen solvated in bulk tungsten. These observations suggest that the role of the yttrium may be to remove the oxygen that may inhibit Cr diffusion to the surface from the tungsten grains. This observation is in accordance with experimental studies showing that the oxidation resistance in the alloy occurs when the oxygen–yttrium ratio in the alloy is close to the stoichiometric ratio for Y<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101987"},"PeriodicalIF":2.7,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145107771","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 : 2025-09-14DOI: 10.1016/j.nme.2025.101989
Haowen Deng , Shouxi Gu , Qiang Qi , Guang-nan Luo
In liquid Pb-Li tritium breeding blanket, the compatibility between liquid Pb-Li and Reduced-activation-ferritic-martensitic (RAFM) steel structure material is crucial for the successful application of the RAFM steel. Corrosion experiments on CLF-1 steel in static Pb-Li at 550 ℃ were conducted for up to 1200 h to figure out the corrosion behavior and mechanism. After exposure, a subset of specimens was cleaned by mixed acid dissolution, while the remaining samples were cold-mounted in resin for cross-sectional observation. X-ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Energy Dispersive Spectroscopy (EDS), Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Electron Backscatter Diffraction (EBSD) measurements were employed to characterize the corrosion behavior and reveal the corrosion mechanism. With increasing exposure time, corrosion initiates at grain boundaries and spreads laterally, producing boundary ditches, hollows, protrusions, and ultimately pebble‑like grains. It is proposed that corrosion occurs through the dissolution of Fe and Cr, facilitated by the diffusion of Li into the grain boundaries and grains, as well as the penetration and adsorption of Pb.
{"title":"Study on the corrosion behavior of static liquid Pb-16.7Li on the structural material CLF-1 steel","authors":"Haowen Deng , Shouxi Gu , Qiang Qi , Guang-nan Luo","doi":"10.1016/j.nme.2025.101989","DOIUrl":"10.1016/j.nme.2025.101989","url":null,"abstract":"<div><div>In liquid Pb-Li tritium breeding blanket, the compatibility between liquid Pb-Li and Reduced-activation-ferritic-martensitic (RAFM) steel structure material is crucial for the successful application of the RAFM steel. Corrosion experiments on CLF-1 steel in static Pb-Li at 550 ℃ were conducted for up to 1200 h to figure out the corrosion behavior and mechanism. After exposure, a subset of specimens was cleaned by mixed acid dissolution, while the remaining samples were cold-mounted in resin for cross-sectional observation. X-ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Energy Dispersive Spectroscopy (EDS), Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Electron Backscatter Diffraction (EBSD) measurements were employed to characterize the corrosion behavior and reveal the corrosion mechanism. With increasing exposure time, corrosion initiates at grain boundaries and spreads laterally, producing boundary ditches, hollows, protrusions, and ultimately pebble‑like grains. It is proposed that corrosion occurs through the dissolution of Fe and Cr, facilitated by the diffusion of Li into the grain boundaries and grains, as well as the penetration and adsorption of Pb.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101989"},"PeriodicalIF":2.7,"publicationDate":"2025-09-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145107770","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 : 2025-09-10DOI: 10.1016/j.nme.2025.101986
Yiqin Wang , Qingmei Xiao , Yang Liu , Shi Ye , Feng Li , Dongye Zhao
Quantifying deuterium (D) retention in plasma-facing components (PFCs) with minimal material impact is critical for fusion reactor operation. This study employs laser-induced desorption coupled with quadrupole mass spectrometry (LID-QMS) for in situ D-retention analysis on HL-3 graphite tiles. As an auxiliary strategy, laser-induced breakdown spectroscopy (LIBS) is implemented under optimized low-fluence conditions to intermittently evaluate LID-QMS desorption efficiency during operation. Laboratory experiments demonstrate > 80 % deuterium release in the first LID pulse (laser fluence > 570 MW/m2), validated via cross-calibrated QMS measurements; LIBS provides rapid efficiency assessment by correlating D/H spectral results with QMS-resolved H, HD and D2 desorption signals. The integrated LID-QMS-LIBS framework permits: real-time optimization of LID parameters during material analysis, direct efficiency validation without destructive sampling. This methodology is currently being implemented on HL-3 tokamak for in situ wall-D monitoring, demonstrating potential to replace ex situ post-mortem analysis in future fusion devices.
{"title":"Efficiency evaluation of fuel retention diagnostic in first wall by LID-QMS: Based on LIBS","authors":"Yiqin Wang , Qingmei Xiao , Yang Liu , Shi Ye , Feng Li , Dongye Zhao","doi":"10.1016/j.nme.2025.101986","DOIUrl":"10.1016/j.nme.2025.101986","url":null,"abstract":"<div><div>Quantifying deuterium (D) retention in plasma-facing components (PFCs) with minimal material impact is critical for fusion reactor operation. This study employs laser-induced desorption coupled with quadrupole mass spectrometry (LID-QMS) for in situ D-retention analysis on HL-3 graphite tiles. As an auxiliary strategy, laser-induced breakdown spectroscopy (LIBS) is implemented under optimized low-fluence conditions to intermittently evaluate LID-QMS desorption efficiency during operation. Laboratory experiments demonstrate > 80 % deuterium release in the first LID pulse (laser fluence > 570 MW/m<sup>2</sup>), validated via cross-calibrated QMS measurements; LIBS provides rapid efficiency assessment by correlating D/H spectral results with QMS-resolved H, HD and D<sub>2</sub> desorption signals. The integrated LID-QMS-LIBS framework permits: real-time optimization of LID parameters during material analysis, direct efficiency validation without destructive sampling. This methodology is currently being implemented on HL-3 tokamak for in situ wall-D monitoring, demonstrating potential to replace ex situ post-mortem analysis in future fusion devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101986"},"PeriodicalIF":2.7,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145057400","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 : 2025-09-10DOI: 10.1016/j.nme.2025.101985
Peng Shao, Xiaowei Ma, Yunshan Xiong, Aoyu Mo, Haijun Li, Quan-Fu Han, Kun Jie Yang, Yue-Lin Liu
Using first-principles calculations, we have systematically explored the geometric structures, electronic properties, diffusion behavior, and clustering with vacancies for impurity B (boron) in tungsten (W) and molybdenum (Mo). A single B atom prefers to occupy octerhedral interstitial position (oip) rather than tetrahedral interstitial position (tip) in bulk metals. B atoms can be easily captured by vacancies, and a single B atom prefers to occupy an oip next to vacancy center with a capturing energies of −2.67 and −2.51 eV in W and Mo, respectively. As the trapping progresses, at least 6B atoms can be captured by one vacancy, which is therefore regarded as the capturing center of B atoms to form BnV clusters in both metals. For interstitial B and mono-vacancy, the most favorable diffusion pathways are oip → tip → oip and the <111> direction, respectively. At the same temperature, the diffusion coefficients of interstitial B and mono-vacancy in W are about 2–9 orders of magnitude lower than those in Mo, indicating that both interstitial B and mono-vacancy migrate much slower in W than in Mo. On the other hand, since the diffusion coefficient of interstitial B is significantly greater than that of mono-vacancy in both metals, the interstitial B migration is much easier than that of mono-vacancy. We therefore conclude that the BnV formation mechanism can be attributed to that the relatively stable vacancies capture these relatively mobile interstitial B atoms in both metals.
{"title":"Energetics and clustering properties of boron in tungsten and molybdenum: A comparative analysis from first-principles study","authors":"Peng Shao, Xiaowei Ma, Yunshan Xiong, Aoyu Mo, Haijun Li, Quan-Fu Han, Kun Jie Yang, Yue-Lin Liu","doi":"10.1016/j.nme.2025.101985","DOIUrl":"10.1016/j.nme.2025.101985","url":null,"abstract":"<div><div>Using first-principles calculations, we have systematically explored the geometric structures, electronic properties, diffusion behavior, and clustering with vacancies for impurity B (boron) in tungsten (W) and molybdenum (Mo). A single B atom prefers to occupy octerhedral interstitial position (oip) rather than tetrahedral interstitial position (tip) in bulk metals. B atoms can be easily captured by vacancies, and a single B atom prefers to occupy an oip next to vacancy center with a capturing energies of −2.67 and −2.51 eV in W and Mo, respectively. As the trapping progresses, at least 6B atoms can be captured by one vacancy, which is therefore regarded as the capturing center of B atoms to form B<sub>n</sub>V clusters in both metals. For interstitial B and mono-vacancy, the most favorable diffusion pathways are oip → tip → oip and the <111> direction, respectively. At the same temperature, the diffusion coefficients of interstitial B and mono-vacancy in W are about 2–9 orders of magnitude lower than those in Mo, indicating that both interstitial B and mono-vacancy migrate much slower in W than in Mo. On the other hand, since the diffusion coefficient of interstitial B is significantly greater than that of mono-vacancy in both metals, the interstitial B migration is much easier than that of mono-vacancy. We therefore conclude that the B<sub>n</sub>V formation mechanism can be attributed to that the relatively stable vacancies capture these relatively mobile interstitial B atoms in both metals.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101985"},"PeriodicalIF":2.7,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145061572","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 : 2025-09-09DOI: 10.1016/j.nme.2025.101984
Yubo Ma , Wei Li , Jun Hu , Xin Zhang , Yuhong Xu , Guangjiu Lei , Shaofei Geng , Haifeng Liu , Xianqu Wang , Jie Huang , Hai Liu , Jun Cheng , Changjian Tang
This study systematically investigates the effects of the stability and work function of barium (Ba) atoms adsorption on the Mo (110) surface using first-principles density functional (DFT) theory calculations. The results demonstrate that the long-bridge site represents the most stable adsorption configuration for Ba atoms on Mo (110) surface. As the Ba coverage increases, the work function initially decreases sharply and then increases slowly, reaching a minimum value of 2.25 eV at a coverage of 4/16 θ (3.35 × 1014 cm−2), which is markedly lower than the work function of 4.85 eV for the clean Mo (110) surface. This indicates that the adsorption of Ba atoms on the Mo (110) surface substantially reduces the work function. Theoretical analysis reveals a linear correlation between work function variations and dipole moment density changes, with charge redistribution induced by Ba adsorption dominating the total dipole moment modification. These results provide the reference for the research of the Cs-free alternative materials for neutral beam injection systems in fusion research.
{"title":"Investigating the stability and work function effects of Ba atoms adsorption on the Mo (110) surface","authors":"Yubo Ma , Wei Li , Jun Hu , Xin Zhang , Yuhong Xu , Guangjiu Lei , Shaofei Geng , Haifeng Liu , Xianqu Wang , Jie Huang , Hai Liu , Jun Cheng , Changjian Tang","doi":"10.1016/j.nme.2025.101984","DOIUrl":"10.1016/j.nme.2025.101984","url":null,"abstract":"<div><div>This study systematically investigates the effects of the stability and work function of barium (Ba) atoms adsorption on the Mo (110) surface using first-principles density functional (DFT) theory calculations. The results demonstrate that the long-bridge site represents the most stable adsorption configuration for Ba atoms on Mo (110) surface. As the Ba coverage increases, the work function initially decreases sharply and then increases slowly, reaching a minimum value of 2.25 eV at a coverage of 4/16 θ (3.35 × 10<sup>14</sup> cm<sup>−2</sup>), which is markedly lower than the work function of 4.85 eV for the clean Mo (110) surface. This indicates that the adsorption of Ba atoms on the Mo (110) surface substantially reduces the work function. Theoretical analysis reveals a linear correlation between work function variations and dipole moment density changes, with charge redistribution induced by Ba adsorption dominating the total dipole moment modification. These results provide the reference for the research of the Cs-free alternative materials for neutral beam injection systems in fusion research.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101984"},"PeriodicalIF":2.7,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145050308","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 : 2025-09-07DOI: 10.1016/j.nme.2025.101983
Jannik Tweer , Thomas Derra , Daniel Dorow-Gerspach , Mauricio Gago , Sascha Gierlings , Stefan Gräfe , Gerald Pintsuk , Marius Wirtz , Christian Linsmeier , Thomas Bergs , Ghaleb Natour
The severe environment and loads acting on plasma-facing components (PFCs) of future fusion power plants cause inevitable erosion of their armor. In situ regeneration of tungsten (W) armored PFCs by local deposition of material would open up the possibility of damage healing and compensation of eroded material. The wire-based laser metal deposition (LMD-w) process fulfils the necessary requirements for use in the reactor vessel. Process development for the deposition of W on W substrate has already been carried out and it has been proven that thermal induced damage in the PFM can be healed this way. In this study, W armored PFCs were coated using LMD-w and tested under fusion–relevant thermal loads in the electron beam facility JUDITH 2. One respectively two stacked layers, each ∼0.65 mm in height, were applied on the top surfaces of the W double tiles with surface areas of 28 × 12 mm2 respectively, which are the characteristic dimensions for the plasma-facing surface of monoblocks. Some of the coated surfaces were also smoothed by laser remelting. In the electron beam facility JUDITH 2, the test components were exposed to steady state as well as combined steady state and transient thermal loads that are expected in the divertor area of the future DEMOnstration power plant. The coatings were tested with cyclic (200 and 1000 cycles) steady state thermal loading in the form of surface temperatures equivalent to heat fluxes on monoblock components of 10 MW m−2 (∼1000 °C) and 15 MW m−2 (∼1500 °C). To determine the performance of LMD-w layers under thermal loads that are expected during exposure to edged localized modes (ELMs), some layers were subjected to combined steady state and transient loading scenarios (0.13–0.55 GW m−2, 103 to 105 pulses of 0.48 ms, 200–700 °C base-temperature). The temperature data obtained from the HHF experiments was processed and analyzed. Profile measurements on the coated surfaces before and after the high heat flux (HHF) exposure were used to investigate the influence of thermal stress on the deposited layers. Furthermore, cross-sectional micrographs of the test components were prepared and analyzed.
{"title":"High heat flux testing of wire-based laser metal deposition coated plasma-facing components","authors":"Jannik Tweer , Thomas Derra , Daniel Dorow-Gerspach , Mauricio Gago , Sascha Gierlings , Stefan Gräfe , Gerald Pintsuk , Marius Wirtz , Christian Linsmeier , Thomas Bergs , Ghaleb Natour","doi":"10.1016/j.nme.2025.101983","DOIUrl":"10.1016/j.nme.2025.101983","url":null,"abstract":"<div><div>The severe environment and loads acting on plasma-facing components (PFCs) of future fusion power plants cause inevitable erosion of their armor. In situ regeneration of tungsten (W) armored PFCs by local deposition of material would open up the possibility of damage healing and compensation of eroded material. The wire-based laser metal deposition (LMD-w) process fulfils the necessary requirements for use in the reactor vessel. Process development for the deposition of W on W substrate has already been carried out and it has been proven that thermal induced damage in the PFM can be healed this way. In this study, W armored PFCs were coated using LMD-w and tested under fusion–relevant thermal loads in the electron beam facility JUDITH 2. One respectively two stacked layers, each <strong>∼</strong>0.65 mm in height, were applied on the top surfaces of the W double tiles with surface areas of 28 × 12 mm<sup>2</sup> respectively, which are the characteristic dimensions for the plasma-facing surface of monoblocks. Some of the coated surfaces were also smoothed by laser remelting. In the electron beam facility JUDITH 2, the test components were exposed to steady state as well as combined steady state and transient thermal loads that are expected in the divertor area of the future DEMOnstration power plant. The coatings were tested with cyclic (200 and 1000 cycles) steady state thermal loading in the form of surface temperatures equivalent to heat fluxes on monoblock components of 10 MW m<sup>−2</sup> (∼1000 °C) and 15 MW m<sup>−2</sup> (∼1500 °C). To determine the performance of LMD-w layers under thermal loads that are expected during exposure to edged localized modes (ELMs), some layers were subjected to combined steady state and transient loading scenarios (0.13–0.55 GW m<sup>−2</sup>, 10<sup>3</sup> to 10<sup>5</sup> pulses of 0.48 ms, 200–700 °C base-temperature). The temperature data obtained from the HHF experiments was processed and analyzed. Profile measurements on the coated surfaces before and after the high heat flux (HHF) exposure were used to investigate the influence of thermal stress on the deposited layers. Furthermore, cross-sectional micrographs of the test components were prepared and analyzed.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101983"},"PeriodicalIF":2.7,"publicationDate":"2025-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145050311","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 : 2025-09-05DOI: 10.1016/j.nme.2025.101982
R.D. Kolasinski , J.D. Coburn , D.D. Truong , J.G. Watkins , T. Abrams , Z.Z. Fang , R. Hood , R.E. Nygren , A.W. Leonard , J. Ren , D.L. Rudakov , J. Sugar , C. Tsui , H.Q. Wang , J.A. Whaley , I. Bykov , A. Cruz , F. Glass , J. Herfindal , C. Lasnier , W. York
In this study, we investigated the effects of combined intense particle and heat flux exposure on advanced tungsten plasma-facing materials within the DIII-D fusion facility. Our test matrix included two types of dispersoid-strengthened tungsten (containing either 100 nm diameter TiO2 or Ni particles), along with high-purity polycrystalline tungsten as a reference. This experiment relied on a sample geometry angled at 15° relative to the divertor surface, thereby allowing the surfaces to intercept steady-state perpendicular heat fluxes () ranging from 10.1 to 19.6 MW/m2. During each shot, the samples were exposed to 42 Hz edge-localized modes (ELMs), allowing us to test the material response to transient heating. We correlated the exposure conditions with extensive post-test surface composition analysis and microscopy to determine how the plasma modified each surface. The angled specimens closest to the strike point received the highest combined heat and particle flux and melted midway through the experiment. EBSD analysis revealed they were completely recrystallized throughout, with an average grain size >100 µm. On the other hand, the specimens that received a lower steady state heat flux survived with more superficial surface damage. Whereas the high-purity polycrystalline tungsten exhibited a higher surface roughness, the dispersoid-strengthened material exhibited more extensive shallow inter-granular cracking. In addition, the surface was depleted of dispersoids following plasma exposure, possibly because of evaporation and/or sputtering. The results described here provide insights into the performance of these materials in a fusion environment which can guide further optimization for use in long-pulse devices.
{"title":"Recrystallization, cracking, and erosion of dispersoid-strengthened tungsten materials during exposure to divertor plasmas","authors":"R.D. Kolasinski , J.D. Coburn , D.D. Truong , J.G. Watkins , T. Abrams , Z.Z. Fang , R. Hood , R.E. Nygren , A.W. Leonard , J. Ren , D.L. Rudakov , J. Sugar , C. Tsui , H.Q. Wang , J.A. Whaley , I. Bykov , A. Cruz , F. Glass , J. Herfindal , C. Lasnier , W. York","doi":"10.1016/j.nme.2025.101982","DOIUrl":"10.1016/j.nme.2025.101982","url":null,"abstract":"<div><div>In this study, we investigated the effects of combined intense particle and heat flux exposure on advanced tungsten plasma-facing materials within the DIII-D fusion facility. Our test matrix included two types of dispersoid-strengthened tungsten (containing either 100 nm diameter TiO<sub>2</sub> or Ni particles), along with high-purity polycrystalline tungsten as a reference. This experiment relied on a sample geometry angled at 15° relative to the divertor surface, thereby allowing the surfaces to intercept steady-state perpendicular heat fluxes (<span><math><mrow><msub><mi>q</mi><mo>⊥</mo></msub></mrow></math></span>) ranging from 10.1 to 19.6 MW/m<sup>2</sup>. During each shot, the samples were exposed to 42 Hz edge-localized modes (ELMs), allowing us to test the material response to transient heating. We correlated the exposure conditions with extensive post-test surface composition analysis and microscopy to determine how the plasma modified each surface. The angled specimens closest to the strike point received the highest combined heat and particle flux and melted midway through the experiment. EBSD analysis revealed they were completely recrystallized throughout, with an average grain size >100 µm. On the other hand, the specimens that received a lower steady state heat flux survived with more superficial surface damage. Whereas the high-purity polycrystalline tungsten exhibited a higher surface roughness, the dispersoid-strengthened material exhibited more extensive shallow inter-granular cracking. In addition, the surface was depleted of dispersoids following plasma exposure, possibly because of evaporation and/or sputtering. The results described here provide insights into the performance of these materials in a fusion environment which can guide further optimization for use in long-pulse devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 101982"},"PeriodicalIF":2.7,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145057401","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 : 2025-09-05DOI: 10.1016/j.nme.2025.101981
S. Markelj , T. Schwarz-Selinger , A. Šestan , J. Zavašnik , M. Kelemen
The effect of near-surface helium (He) on deuterium (D) retention and uptake into the bulk of tungsten (W) was investigated. To quantify the He influence on D uptake, He was implanted close to the surface with 3 keV energy at different fluences and different temperatures. 20 MeV W irradiation was performed at room temperature after He implantation to create defects within the first 2.3 µm. Samples were then exposed to a low flux, low energy (300 eV/D) D ion beam at 450 K. The defects created by W ions trap penetrating D and make it hence possible to quantify D transport into depth below the He layer using 3He nuclear reaction analysis. Elastic recoil detection analysis enabled us to measure the D and He concentration depth profiles near the surface. Results show that D gets preferentially retained where He is implanted with D concentrations up to 10 at.%. At the same time D uptake beyond the He zone is reduced by a factor of 15 compared to a He-free W sample.
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Pub Date : 2025-09-01DOI: 10.1016/j.nme.2025.101968
Jasper Ristkok , Salvatore Almaviva , Jari Likonen , Juuso Karhunen , Indrek Jõgi , Peeter Paris , Shweta Soni , Pavel Veis , Sahithya Atikukke , Jelena Butikova , Rongxing Yi , Ionut Jepu , Pawel Gasior , Corneliu Porosnicu , Mihaela Bojan , Bianca Solomonea , Sebastijan Brezinsek
Laser-induced breakdown spectroscopy (LIBS) is a method for elemental composition analysis that has been proposed for fusion reactor safety diagnostics. A significant milestone in this development was the LIBS campaign conducted in 2024 at the Joint European Torus (JET), using a prototype LIBS enclosure, deployed with the MASCOT tele-manipulation arm. The work presented here prepared for the JET campaign by testing the LIBS enclosure.
Experiments were conducted at VTT Technical Research Centre of Finland, analyzing JET wall samples from the 2011–2016 ILW1–3 fusion campaigns, primarily from the divertor. The focus was on the analysis of co-deposited layers on the plasma-facing components containing hydrogen isotopes and elements from bulk layers: Be, W, Mo, CFC, and Inconel. Measurements were performed under atmospheric pressure air with an argon flow.
Optimal experimental conditions for the use of an Echelle spectrometer in subsequent JET LIBS campaign were identified, and the depth profiles of the surface layers are presented. The LIBS depth profiles defined distinct material layers. Ablating through the co-deposited layers required 1–870 laser shots (∼0.1–90 µm) on samples from different locations, with typical variations of 10–40 % on the same sample and the largest variation spanning 15–480 shots (∼1.5–50 µm).
The LIBS, Secondary Ion Mass Spectrometry (SIMS), and optical profilometry results showed good qualitative agreement. The ablation rate was ∼30–50 nm/shot for the W layers, ∼100–140 nm/shot for bulk Be limiters, and intermediate for the co-deposited layers. The insights gained in this study supported the preparation of the JET LIBS campaign.
{"title":"Preparing LIBS for in-situ measurements in JET tokamak: system overview and co-deposited layer thicknesses","authors":"Jasper Ristkok , Salvatore Almaviva , Jari Likonen , Juuso Karhunen , Indrek Jõgi , Peeter Paris , Shweta Soni , Pavel Veis , Sahithya Atikukke , Jelena Butikova , Rongxing Yi , Ionut Jepu , Pawel Gasior , Corneliu Porosnicu , Mihaela Bojan , Bianca Solomonea , Sebastijan Brezinsek","doi":"10.1016/j.nme.2025.101968","DOIUrl":"10.1016/j.nme.2025.101968","url":null,"abstract":"<div><div>Laser-induced breakdown spectroscopy (LIBS) is a method for elemental composition analysis that has been proposed for fusion reactor safety diagnostics. A significant milestone in this development was the LIBS campaign conducted in 2024 at the Joint European Torus (JET), using a prototype LIBS enclosure, deployed with the MASCOT tele-manipulation arm. The work presented here prepared for the JET campaign by testing the LIBS enclosure.</div><div>Experiments were conducted at VTT Technical Research Centre of Finland, analyzing JET wall samples from the 2011–2016 ILW1–3 fusion campaigns, primarily from the divertor. The focus was on the analysis of co-deposited layers on the plasma-facing components containing hydrogen isotopes and elements from bulk layers: Be, W, Mo, CFC, and Inconel. Measurements were performed under atmospheric pressure air with an argon flow.</div><div>Optimal experimental conditions for the use of an Echelle spectrometer in subsequent JET LIBS campaign were identified, and the depth profiles of the surface layers are presented. The LIBS depth profiles defined distinct material layers. Ablating through the co-deposited layers required 1–870 laser shots (∼0.1–90 µm) on samples from different locations, with typical variations of 10–40 % on the same sample and the largest variation spanning 15–480 shots (∼1.5–50 µm).</div><div>The LIBS, Secondary Ion Mass Spectrometry (SIMS), and optical profilometry results showed good qualitative agreement. The ablation rate was ∼30–50 nm/shot for the W layers, ∼100–140 nm/shot for bulk Be limiters, and intermediate for the co-deposited layers. The insights gained in this study supported the preparation of the JET LIBS campaign.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"44 ","pages":"Article 101968"},"PeriodicalIF":2.7,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145048916","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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