Pub Date : 2025-03-31DOI: 10.1016/j.nme.2025.101927
Anastacia Wright , Benedict Keates , Zhexin Cui , Eric Prestat , Sergio Lozano-Perez , Liberato Volpe
The reduced activation ferritic/martensitic (RAF/M) EUROFER97 is the primary candidate for coolant-facing alloys of the European DEMO reactor. However, EUROFER97 will face harsh environment, and it might experience enhanced degradation due to the synergistic exposure to flowing high temperature water coolant, tritium embrittlement and ionising radiation. A possible solution is to use self-passivating coatings that can interact with the environment to provide corrosion and tritium permeation protection during in service operation. In this study, EUROFER97 was firstly Al-base coated using either a chemical vapour deposition (CVD at 700 °C or 750 °C) or electroplating process (room temperature); and then oxidised in a 2-step process using the same tempering (980 °C) and normalisation (760 °C) temperatures used for EUROFER97. These temperatures were used to assess their suitability for obtaining a stable and protective Al-base oxide without altering the microstructure of EUROFER97. Advanced microscopy characterizations revealed the formation of a ≈0.6 µm α-Al2O3 layer over an intermetallic Al-Fe-rich interdiffusion microstructure of ≈80 µm depth. However, the interdiffusion layer was highly decorated with voids that might act as stress concentrators during in-plant service, thus being detrimental to the material performance. This study compares different Al-base coating techniques and provides a preliminary insight on the selection of oxidation temperatures for EUROFER97, finding that the current 2-step oxidation process needs further optimisation before being industrialised.
{"title":"Oxidation behaviour of CVD and electroplated Al-base coatings for EUROFER97","authors":"Anastacia Wright , Benedict Keates , Zhexin Cui , Eric Prestat , Sergio Lozano-Perez , Liberato Volpe","doi":"10.1016/j.nme.2025.101927","DOIUrl":"10.1016/j.nme.2025.101927","url":null,"abstract":"<div><div>The reduced activation ferritic/martensitic (RAF/M) EUROFER97 is the primary candidate for coolant-facing alloys of the European DEMO reactor. However, EUROFER97 will face harsh environment, and it might experience enhanced degradation due to the synergistic exposure to flowing high temperature water coolant, tritium embrittlement and ionising radiation. A possible solution is to use self-passivating coatings that can interact with the environment to provide corrosion and tritium permeation protection during in service operation. In this study, EUROFER97 was firstly Al-base coated using either a chemical vapour deposition (CVD at 700 °C or 750 °C) or electroplating process (room temperature); and then oxidised in a 2-step process using the same tempering (980 °C) and normalisation (760 °C) temperatures used for EUROFER97. These temperatures were used to assess their suitability for obtaining a stable and protective Al-base oxide without altering the microstructure of EUROFER97. Advanced microscopy characterizations revealed the formation of a ≈0.6 µm α-Al<sub>2</sub>O<sub>3</sub> layer over an intermetallic Al-Fe-rich interdiffusion microstructure of ≈80 µm depth. However, the interdiffusion layer was highly decorated with voids that might act as stress concentrators during in-plant service, thus being detrimental to the material performance. This study compares different Al-base coating techniques and provides a preliminary insight on the selection of oxidation temperatures for EUROFER97, finding that the current 2-step oxidation process needs further optimisation before being industrialised.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101927"},"PeriodicalIF":2.3,"publicationDate":"2025-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143746396","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-28DOI: 10.1016/j.nme.2025.101923
Volker Rohde, Martin Balden, Karl Krieger, Rudolf Neu, ASDEX Upgrade Team
With the switch to tungsten for the ITER plasma facing components, wall conditioning techniques such as boronization are gaining new interest. As AUG uses tungsten plasma facing components as well as boronization, the present results are summarized and reviewed in this paper. In AUG it has been shown how to operate without boronization, but conditioning by B coating through a glow discharge (boronization) is now the standard start-up procedure. The properties of the amorphous boron hydride layers, produced by boronization, depend on the discharge conditions and the residual carbon content. At AUG, chemically active layers are required to getter oxygen and deuterium. After initial conditioning, further boronization is only required for some special scenarios. The lifetime of these subsequent coatings is about 20–30 discharges, after which a new coating or a boron powder injection is required to refresh the conditioning for these special scenarios. The main effect of boronization is to reduce tungsten sputtering from the main chamber limiter by reducing impurities. As the layers are chemically active, they react with air during venting, producing unstable whitish layers that are slowly vanishing. Specifically, they cannot be studied by scanning electron microscopy or ion beam analysis, because under vacuum conditions they are lost even faster. The remaining layers are long-term stable, but contain on average only 4.1 % of the boron injected during the boronizations. This remobilisation reduces the deuterium inventory in the layers to about 0.1 % of the input during the campaigns.
{"title":"Boronization with tungsten plasma-facing surfaces in ASDEX Upgrade","authors":"Volker Rohde, Martin Balden, Karl Krieger, Rudolf Neu, ASDEX Upgrade Team","doi":"10.1016/j.nme.2025.101923","DOIUrl":"10.1016/j.nme.2025.101923","url":null,"abstract":"<div><div>With the switch to tungsten for the ITER plasma facing components, wall conditioning techniques such as boronization are gaining new interest. As AUG uses tungsten plasma facing components as well as boronization, the present results are summarized and reviewed in this paper. In AUG it has been shown how to operate without boronization, but conditioning by B coating through a glow discharge (boronization) is now the standard start-up procedure. The properties of the amorphous boron hydride layers, produced by boronization, depend on the discharge conditions and the residual carbon content. At AUG, chemically active layers are required to getter oxygen and deuterium. After initial conditioning, further boronization is only required for some special scenarios. The lifetime of these subsequent coatings is about 20–30 discharges, after which a new coating or a boron powder injection is required to refresh the conditioning for these special scenarios. The main effect of boronization is to reduce tungsten sputtering from the main chamber limiter by reducing impurities. As the layers are chemically active, they react with air during venting, producing unstable whitish layers that are slowly vanishing. Specifically, they cannot be studied by scanning electron microscopy or ion beam analysis, because under vacuum conditions they are lost even faster. The remaining layers are long-term stable, but contain on average only 4.1 % of the boron injected during the boronizations. This remobilisation reduces the deuterium inventory in the layers to about 0.1 % of the input during the campaigns.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101923"},"PeriodicalIF":2.3,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143746397","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-25DOI: 10.1016/j.nme.2025.101916
M. Bernert , T.O.S.J. Bosman , T. Lunt , O. Pan , B. Sieglin , U. Stroth , A. Kallenbach , S. Wiesen , M. Wischmeier , G. Birkenmeier , M. Cavedon , B. Lipschultz , C. Lowry , N. Fedorczak , P. Fox , M. Lennholm , H. Sun , P. Jacquet , K. Kirov , N. Vianello , F. Reimold
Power exhaust is a crucial issue for future fusion reactors. Divertor detachment and the required power dissipation fractions of about 95% are foreseen to be achieved by impurity seeding. In a tokamak, at high seeding levels the radiation often concentrates in a small region inside the confined plasma near the X-point. In early observations the so-called X-point radiator (XPR) often led to back-transitions to L-mode or disruptions. In metal tokamaks or with higher available heating power, these regimes can be stabilized and are now established on AUG, JET, TCV, KSTAR and WEST.
The XPR is a cold, dense plasma inside the confined region in the vicinity of the X-point, that breaks the paradigm of poloidal symmetry of density and temperature on closed flux surfaces. On AUG, the poloidal extent of the XPR is a few centimeters and it is observed up to 15 above the X-point. The long connection length in this region and the access of neutral particles from the divertor region facilitate the creation of the XPR, as predicted by an analytical model. Numerical simulations with SOLPS-ITER match the observations at AUG and TCV and allow predictions towards a power plant, where a lower impurity concentration is required to trigger an XPR. Since the XPR greatly reduces power and particle fluxes to the targets, simpler and more efficient divertor concepts, such as the compact radiative divertor, can be envisaged for future devices. A scenario with an XPR, however, comes at the cost of an increased impurity concentration and a potential reduction in confinement, which has to be further quantified.
The XPR location can be well detected by various diagnostics, enabling responsive real-time control, even through large transients like an LH transition. The active control helped to access a new regime of ELM suppression at AUG, which is now also observed at TCV and JET.
The observation of the XPR on multiple tokamaks, the demonstration of its active control, and the emergence of theoretical models that scale favourably towards fusion reactors have opened up a new phase of advanced power exhaust research.
功率耗尽是未来聚变反应堆的一个关键问题。可以预见，通过添加杂质，可以实现掺混器分离和所需的约 95% 的功率耗散分数。在托卡马克中，当种子水平较高时，辐射通常会集中在封闭等离子体内部靠近 X 点的一小块区域。在早期的观测中，所谓的 X 点辐射器（XPR）经常导致向 L 模式的反向转换或中断。在金属托卡马克或更高的可用加热功率下，这些状态可以被稳定下来，现在已经在AUG、JET、TCV、KSTAR和WEST上建立起来。XPR是X点附近约束区域内的冷致密等离子体，它打破了封闭通量面上密度和温度的极对称范式。在 AUG 上，XPR 的极坐标范围为几厘米，在 X 点上方 15 c m 处可以观察到。正如分析模型所预测的那样，该区域的长连接长度和中性粒子从分流区的进入促进了 XPR 的产生。利用 SOLPS-ITER 进行的数值模拟与在 AUG 和 TCV 的观测结果相吻合，并可对发电厂进行预测，在发电厂，需要较低的杂质浓度才能触发 XPR。由于 XPR 大大降低了目标的功率和粒子通量，因此可以为未来的装置设想更简单、更高效的分流器概念，如紧凑型辐射分流器。不过，使用 XPR 的方案需要以增加杂质浓度和可能降低约束性为代价，这一点还需要进一步量化。XPR 的位置可以通过各种诊断仪很好地检测到，从而实现响应式实时控制，即使是在 LH 转变等大的瞬态情况下也是如此。在多个托卡马克上观测到的 XPR、其主动控制的演示以及有利于聚变反应堆扩展的理论模型的出现，开启了先进功率排气研究的新阶段。
{"title":"X-point radiation: From discovery to potential application in a future reactor","authors":"M. Bernert , T.O.S.J. Bosman , T. Lunt , O. Pan , B. Sieglin , U. Stroth , A. Kallenbach , S. Wiesen , M. Wischmeier , G. Birkenmeier , M. Cavedon , B. Lipschultz , C. Lowry , N. Fedorczak , P. Fox , M. Lennholm , H. Sun , P. Jacquet , K. Kirov , N. Vianello , F. Reimold","doi":"10.1016/j.nme.2025.101916","DOIUrl":"10.1016/j.nme.2025.101916","url":null,"abstract":"<div><div>Power exhaust is a crucial issue for future fusion reactors. Divertor detachment and the required power dissipation fractions of about 95% are foreseen to be achieved by impurity seeding. In a tokamak, at high seeding levels the radiation often concentrates in a small region inside the confined plasma near the X-point. In early observations the so-called X-point radiator (XPR) often led to back-transitions to L-mode or disruptions. In metal tokamaks or with higher available heating power, these regimes can be stabilized and are now established on AUG, JET, TCV, KSTAR and WEST.</div><div>The XPR is a cold, dense plasma inside the confined region in the vicinity of the X-point, that breaks the paradigm of poloidal symmetry of density and temperature on closed flux surfaces. On AUG, the poloidal extent of the XPR is a few centimeters and it is observed up to 15<!--> <span><math><mi>c</mi></math></span> <span><math><mi>m</mi></math></span> above the X-point. The long connection length in this region and the access of neutral particles from the divertor region facilitate the creation of the XPR, as predicted by an analytical model. Numerical simulations with SOLPS-ITER match the observations at AUG and TCV and allow predictions towards a power plant, where a lower impurity concentration is required to trigger an XPR. Since the XPR greatly reduces power and particle fluxes to the targets, simpler and more efficient divertor concepts, such as the compact radiative divertor, can be envisaged for future devices. A scenario with an XPR, however, comes at the cost of an increased impurity concentration and a potential reduction in confinement, which has to be further quantified.</div><div>The XPR location can be well detected by various diagnostics, enabling responsive real-time control, even through large transients like an LH transition. The active control helped to access a new regime of ELM suppression at AUG, which is now also observed at TCV and JET.</div><div>The observation of the XPR on multiple tokamaks, the demonstration of its active control, and the emergence of theoretical models that scale favourably towards fusion reactors have opened up a new phase of advanced power exhaust research.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101916"},"PeriodicalIF":2.3,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143739927","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-24DOI: 10.1016/j.nme.2025.101924
Qian Wang , Liang Chen , Chaoping Liang
First-principles calculations were utilized to explore the effects of Zr, Nb, Ta, and W on the solubility, diffusivity, and permeability of hydrogen in TiC. It is found that the trigonal site of H in TiC, which is encircled by three Ti atoms lying a {111} plane, has been identified as thermodynamically more stable. The addition of alloying elements will destroy the equilibrium distribution of the Ti-H interaction and reduce the structural stability of interstitial H. Furthermore, the solubility, diffusivity, and permeability of H in Ti32C32 and Ti31MC32 (M = Zr, Nb, Ta, and W) at finite temperatures are derived, and the results are discussed and compared with similar experiments in the literature. Drawing upon the influence of elemental alloying on H retention, we deeply understand experimental observations and provide experimentalists with a potentially valuable recommendation for controlling hydrogen permeability.
{"title":"Effects of alloying elements on the solubility, diffusivity, and permeability of hydrogen in TiC from first-principles calculation","authors":"Qian Wang , Liang Chen , Chaoping Liang","doi":"10.1016/j.nme.2025.101924","DOIUrl":"10.1016/j.nme.2025.101924","url":null,"abstract":"<div><div>First-principles calculations were utilized to explore the effects of Zr, Nb, Ta, and W on the solubility, diffusivity, and permeability of hydrogen in TiC. It is found that the trigonal site of H in TiC, which is encircled by three Ti atoms lying a {111} plane, has been identified as thermodynamically more stable. The addition of alloying elements will destroy the equilibrium distribution of the Ti-H interaction and reduce the structural stability of interstitial H. Furthermore, the solubility, diffusivity, and permeability of H in Ti<sub>32</sub>C<sub>32</sub> and Ti<sub>31</sub>MC<sub>32</sub> (M = Zr, Nb, Ta, and W) at finite temperatures are derived, and the results are discussed and compared with similar experiments in the literature. Drawing upon the influence of elemental alloying on H retention, we deeply understand experimental observations and provide experimentalists with a potentially valuable recommendation for controlling hydrogen permeability.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101924"},"PeriodicalIF":2.3,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143705325","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-20DOI: 10.1016/j.nme.2025.101922
Yongkui Wang , Xiaochen Huang , Changpeng Lv , Zhong Wu , Jinlong Ge , Jianguo Ma , Xueqi Zhang
This study investigates the oxidative behavior of pure tungsten and a W-0.5 wt% ZrC alloy (WZC05), was examined. Both materials were fabricated via sintering and hot rolling. The alloys were exposed to air at 700 °C for 20 h and to air at 900 °C for 8 h to analyze surface morphology, oxidation resistance, and oxidation behavior of pure tungsten and WZC05. WZC05 exhibited a significantly smaller grain size than pure tungsten. At 700 °C, the “positive effect of grain size,” caused a higher oxidation rate in pure tungsten compared with that in WZC05. After air oxidation for 20 h at 700 °C, the thickness of the oxide layer on pure tungsten reached 320 μm, whereas that on WZC05 was only 202 μm. At 900 °C, the dense oxide layer was destroyed owing to the formation of volatile WO3, resulting in similar oxidation rates for pure tungsten and WZC05. Scanning electron micrographs revealed that the oxidized tungsten materials had a layered structure.
{"title":"Oxidation behavior of pure W and W-0.5 wt% ZrC at 700 °C and 900 °C","authors":"Yongkui Wang , Xiaochen Huang , Changpeng Lv , Zhong Wu , Jinlong Ge , Jianguo Ma , Xueqi Zhang","doi":"10.1016/j.nme.2025.101922","DOIUrl":"10.1016/j.nme.2025.101922","url":null,"abstract":"<div><div>This study investigates the oxidative behavior of pure tungsten and a W-0.5 wt% ZrC alloy (WZC05), was examined. Both materials were fabricated via sintering and hot rolling. The alloys were exposed to air at 700 °C for 20 h and to air at 900 °C for 8 h to analyze surface morphology, oxidation resistance, and oxidation behavior of pure tungsten and WZC05. WZC05 exhibited a significantly smaller grain size than pure tungsten. At 700 °C, the “positive effect of grain size,” caused a higher oxidation rate in pure tungsten compared with that in WZC05. After air oxidation for 20 h at 700 °C, the thickness of the oxide layer on pure tungsten reached 320 μm, whereas that on WZC05 was only 202 μm. At 900 °C, the dense oxide layer was destroyed owing to the formation of volatile WO<sub>3</sub>, resulting in similar oxidation rates for pure tungsten and WZC05. Scanning electron micrographs revealed that the oxidized tungsten materials had a layered structure.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101922"},"PeriodicalIF":2.3,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143704095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-15DOI: 10.1016/j.nme.2025.101920
T. Hernández , M. Roldán , F.J. Sánchez , N. García-Rodríguez , J. Patiño , J. Navas , E. Carella , L. Serrador , I. Rueda , E. Abad , R. San Martín
The weldability of Eurofer steel and the behavior of weld beads under relevant fusion conditions are critical factors to assess for future designs, specifically in the context of the Water-Cooled Lithium Lead (WCLL) blanket concept. Here, the area corresponding to the breeder blanket will be in direct contact with the PbLi eutectic alloy, which will flow continuously at temperatures between 400 °C and 550 °C. In this study, electron beam welding was performed on 1.4 mm thick Eurofer plates. These welded samples were subsequently exposed to flowing PbLi in the CiCLo loop at CIEMAT, maintained at 450 °C for 1000 h.
Prior to corrosion testing, the microstructure of the samples was characterized to evaluate the fusion zone width, bonding coefficient, and the presence of any defects or brittle joints. Acicular delta ferrite was observed, attributed to the high temperatures during welding and rapid cooling afterward. The potential for improved corrosion resistance through normalization treatment post-welding was also assessed. Following PbLi exposure, the specimens underwent the same microstructural characterization.
The results showed a significant microstructural change in the HAZ, which caused an increase in hardness in the HAZ; this result occurred in both the as-welded and heat-treated materials. It should be noted that non-negligible amounts of lithium were found in the HAZ, which is evidence of the diffusion of lithium from the molten eutectic into the interior of the material.
{"title":"Electron beam welding behavior in Eurofer subjected to the PbLi eutectic","authors":"T. Hernández , M. Roldán , F.J. Sánchez , N. García-Rodríguez , J. Patiño , J. Navas , E. Carella , L. Serrador , I. Rueda , E. Abad , R. San Martín","doi":"10.1016/j.nme.2025.101920","DOIUrl":"10.1016/j.nme.2025.101920","url":null,"abstract":"<div><div>The weldability of Eurofer steel and the behavior of weld beads under relevant fusion conditions are critical factors to assess for future designs, specifically in the context of the Water-Cooled Lithium Lead (WCLL) blanket concept. Here, the area corresponding to the breeder blanket will be in direct contact with the PbLi eutectic alloy, which will flow continuously at temperatures between 400 °C and 550 °C. In this study, electron beam welding was performed on 1.4 mm thick Eurofer plates. These welded samples were subsequently exposed to flowing PbLi in the CiCLo loop at CIEMAT, maintained at 450 °C for 1000 h.</div><div>Prior to corrosion testing, the microstructure of the samples was characterized to evaluate the fusion zone width, bonding coefficient, and the presence of any defects or brittle joints. Acicular delta ferrite was observed, attributed to the high temperatures during welding and rapid cooling afterward. The potential for improved corrosion resistance through normalization treatment post-welding was also assessed. Following PbLi exposure, the specimens underwent the same microstructural characterization.</div><div>The results showed a significant microstructural change in the HAZ, which caused an increase in hardness in the HAZ; this result occurred in both the as-welded and heat-treated materials. It should be noted that non-negligible amounts of lithium were found in the HAZ, which is evidence of the diffusion of lithium from the molten eutectic into the interior of the material.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101920"},"PeriodicalIF":2.3,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143679241","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-13DOI: 10.1016/j.nme.2025.101919
Yi Li , Dahuan Zhu , Chunyu He , Zongxiao Guo , Yang Wang , Chuannan Xuan , Baoguo Wang , Junling Chen , EAST Team
In fusion devices, plasma-facing components (PFCs) play a critical role in withstanding severe thermal conditions resulting from cyclic high heat flux (HHF) loads. The International Thermonuclear Experimental Reactor (ITER) and next-generation fusion devices are expected to employ actively cooled tungsten/copper (W/Cu) monoblocks as divertor targets due to their excellent heat removal capabilities. Although ITER-like monoblocks utilize a soft Cu interlayer to alleviate stress, interface fatigue cracking still occurs under cyclic HHF loads. The issue of interface bonding between the W armor and heat sink has been a limiting factor for the long-term stable operation and structural integrity of these monoblocks. Functionally graded materials (FGMs) are regarded as an effective approach to address severe local stress concentration at the bonding interface. The number of layers, composition distribution, and thickness of the FGM layers are analyzed by evaluating the stress and strain after the loading and cooling phases in finite element simulations. The simulation results indicate that the graded interlayer can significantly reduce stress concentration at the interface, and a two-layer FGM (25 vol% and 66.7 vol% W) with each layer 0.6 mm thick can greatly mitigate both stress and strain. Such results provide important guidance for the development of graded W/Cu monoblocks for fusion applications.
{"title":"Thermomechanical analysis for the theoretical optimization of W/Cu monoblocks with functionally graded interlayer","authors":"Yi Li , Dahuan Zhu , Chunyu He , Zongxiao Guo , Yang Wang , Chuannan Xuan , Baoguo Wang , Junling Chen , EAST Team","doi":"10.1016/j.nme.2025.101919","DOIUrl":"10.1016/j.nme.2025.101919","url":null,"abstract":"<div><div>In fusion devices, plasma-facing components (PFCs) play a critical role in withstanding severe thermal conditions resulting from cyclic high heat flux (HHF) loads. The International Thermonuclear Experimental Reactor (ITER) and next-generation fusion devices are expected to employ actively cooled tungsten/copper (W/Cu) monoblocks as divertor targets due to their excellent heat removal capabilities. Although ITER-like monoblocks utilize a soft Cu interlayer to alleviate stress, interface fatigue cracking still occurs under cyclic HHF loads. The issue of interface bonding between the W armor and heat sink has been a limiting factor for the long-term stable operation and structural integrity of these monoblocks. Functionally graded materials (FGMs) are regarded as an effective approach to address severe local stress concentration at the bonding interface. The number of layers, composition distribution, and thickness of the FGM layers are analyzed by evaluating the stress and strain after the loading and cooling phases in finite element simulations. The simulation results indicate that the graded interlayer can significantly reduce stress concentration at the interface, and a two-layer FGM (25 vol% and 66.7 vol% W) with each layer 0.6 mm thick can greatly mitigate both stress and strain. Such results provide important guidance for the development of graded W/Cu monoblocks for fusion applications.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101919"},"PeriodicalIF":2.3,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143644865","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-13DOI: 10.1016/j.nme.2025.101921
Hanfeng Song , Chao Qi , Jiaguan Peng , Pengcheng Guo , Junyun Lai , Long Cheng , Yue Yuan , Bo Wang , Guang-Hong Lu
Tungsten (W) is a promising candidate for plasma-facing materials in fusion reactors; however, its application is hindered by challenges such as blistering. This study proposes a laminated microstructure W design, developed by stacking warm-rolled W foils with thicknesses of 0.05 mm and 0.1 mm. The plasma-exposed surface exhibited a strong preferential [110] orientation, nanoscale grains, and grain boundaries oriented perpendicular to the surface, in addition to interlayer gaps between the foils. Laminated samples, composed of laminated microstructure W and 2.5 mm thick warm-rolled and recrystallized W bulks, were fabricated and exposed to deuterium plasma at a flux of 3 × 1020 ions m-2s−1, with fluences of 1 × 1025 ions m−2 and 5 × 1025 ions m−2. The results demonstrated that the laminated microstructure W exhibits superior resistance to blistering. Furthermore, laminated W foils were successfully brazed onto a Cu substrate, validating the feasibility of manufacturing laminated W plasma-facing component (PFC). These findings indicate that laminated W-based PFC represent a promising design strategy for improving the irradiation tolerance of PFC under fusion reactor conditions.
钨（W）是核聚变反应堆等离子体面材料的理想候选材料，但其应用却受到起泡等难题的阻碍。本研究提出了一种层状微结构钨设计，它是通过堆叠厚度为 0.05 毫米和 0.1 毫米的热轧钨箔而开发的。等离子体暴露的表面表现出强烈的优先[110]取向、纳米级晶粒和垂直于表面的晶界，此外，箔片之间还存在层间间隙。层叠样品由层叠微结构 W 和 2.5 mm 厚的热轧再结晶 W 块体组成，样品制作完成后暴露在通量为 3 × 1020 离子 m-2s-1 的氘等离子体中，通量分别为 1 × 1025 离子 m-2 和 5 × 1025 离子 m-2。结果表明，层压微结构 W 具有优异的抗起泡能力。此外，层叠 W 箔还成功地钎焊到了铜基板上，验证了制造层叠 W 等离子体面组件 (PFC) 的可行性。这些研究结果表明，层压 W 基 PFC 是一种很有前途的设计策略，可提高 PFC 在聚变反应堆条件下的辐照耐受性。
{"title":"Resistance to deuterium-induced blistering in laminated microstructure tungsten","authors":"Hanfeng Song , Chao Qi , Jiaguan Peng , Pengcheng Guo , Junyun Lai , Long Cheng , Yue Yuan , Bo Wang , Guang-Hong Lu","doi":"10.1016/j.nme.2025.101921","DOIUrl":"10.1016/j.nme.2025.101921","url":null,"abstract":"<div><div>Tungsten (W) is a promising candidate for plasma-facing materials in fusion reactors; however, its application is hindered by challenges such as blistering. This study proposes a laminated microstructure W design, developed by stacking warm-rolled W foils with thicknesses of 0.05 mm and 0.1 mm. The plasma-exposed surface exhibited a strong preferential [110] orientation, nanoscale grains, and grain boundaries oriented perpendicular to the surface, in addition to interlayer gaps between the foils. Laminated samples, composed of laminated microstructure W and 2.5 mm thick warm-rolled and recrystallized W bulks, were fabricated and exposed to deuterium plasma at a flux of 3 × 10<sup>20</sup> ions m<sup>-2</sup>s<sup>−1</sup>, with fluences of 1 × 10<sup>25</sup> ions m<sup>−2</sup> and 5 × 10<sup>25</sup> ions m<sup>−2</sup>. The results demonstrated that the laminated microstructure W exhibits superior resistance to blistering. Furthermore, laminated W foils were successfully brazed onto a Cu substrate, validating the feasibility of manufacturing laminated W plasma-facing component (PFC). These findings indicate that laminated W-based PFC represent a promising design strategy for improving the irradiation tolerance of PFC under fusion reactor conditions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101921"},"PeriodicalIF":2.3,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143629496","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-10DOI: 10.1016/j.nme.2025.101918
Yu Tian , Yu-Ping Xu , Chang Xu , Wu-Qingliang Peng , An-Bang Wu , Jian-Ping Qu , Fang-Yong Du , Peng-Fei Zi , Qiang Li , Wan-Jing Wang , Hai-Shan Zhou , Guang-Nan Luo
The supporting legs at the back of the ITER-like W/Cu monoblock in the upper divertor of the Experimental Advanced Superconducting Tokamak (EAST) were found fractured due to the influence of electromagnetic forces, vertical displacement events (VDEs) and its own weight. This study developed an electron beam brazing (EBB) technique for the reimplantation of supporting legs. Defect-free bonding between Inconel 625 (IN625) alloy and pure tungsten was obtained with the tensile strength exceeding 110 MPa. Microstructural analysis revealed a columnar grain structure on the tungsten side and finer grains on the IN625 side. Nanoindentation tests indicated that the brazed interface exhibited a lower average stiffness of 1660 μN/nm and a lower average hardness of 1.48 GPa, ensuring enhanced plasticity and improved load-bearing capacity. This work provides a novel proposal to reimplant supporting legs on divertor components by EBB in terms of achieving microstructural refinement and high strength.
{"title":"Reimplantation of supporting legs on EAST divertor by electron beam brazing","authors":"Yu Tian , Yu-Ping Xu , Chang Xu , Wu-Qingliang Peng , An-Bang Wu , Jian-Ping Qu , Fang-Yong Du , Peng-Fei Zi , Qiang Li , Wan-Jing Wang , Hai-Shan Zhou , Guang-Nan Luo","doi":"10.1016/j.nme.2025.101918","DOIUrl":"10.1016/j.nme.2025.101918","url":null,"abstract":"<div><div>The supporting legs at the back of the ITER-like W/Cu monoblock in the upper divertor of the Experimental Advanced Superconducting Tokamak (EAST) were found fractured due to the influence of electromagnetic forces, vertical displacement events (VDEs) and its own weight. This study developed an electron beam brazing (EBB) technique for the reimplantation of supporting legs. Defect-free bonding between Inconel 625 (IN625) alloy and pure tungsten was obtained with the tensile strength exceeding 110 MPa. Microstructural analysis revealed a columnar grain structure on the tungsten side and finer grains on the IN625 side. Nanoindentation tests indicated that the brazed interface exhibited a lower average stiffness of 1660 μN/nm and a lower average hardness of 1.48 GPa, ensuring enhanced plasticity and improved load-bearing capacity. This work provides a novel proposal to reimplant supporting legs on divertor components by EBB in terms of achieving microstructural refinement and high strength.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101918"},"PeriodicalIF":2.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143600967","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-10DOI: 10.1016/j.nme.2025.101912
Y. Damizia , S. Elmore , P. Ryan , S. Allan , F. Federici , N. Osborne , J.W. Bradley , MAST-U Team
This study presents the first ion temperature () measurements from the MAST-U divertor using a Retarding Field Energy Analyzer (RFEA). Embedded within the flat tile of the closed divertor chamber, the RFEA captures profiles across various plasma scenarios, including transitions to the Super-X configuration. Measurements were conducted under steady-state and transient plasma conditions characterized by a plasma current () of 750 kA, electron density () between and , and Neutral Beam Injection (NBI) power ranging from 3.0 MW to 3.2 MW. The ion temperatures, peaking at approximately 17 eV in steady state, were compared with electron temperatures () obtained from Langmuir probes (LP) at identical radial positions. Preliminary findings reveal a ratio ranging from 1 to 2.2. Additionally, high temporal resolution measurements () captured the dynamics of Edge Localized Modes (ELMs), showing peaks at 16 ±1.8 eV during ELM events, nearly three times higher than inter-ELM temperatures.
{"title":"First ion temperature measurements in the MAST-U divertor via Retarding Field Energy Analyzer","authors":"Y. Damizia , S. Elmore , P. Ryan , S. Allan , F. Federici , N. Osborne , J.W. Bradley , MAST-U Team","doi":"10.1016/j.nme.2025.101912","DOIUrl":"10.1016/j.nme.2025.101912","url":null,"abstract":"<div><div>This study presents the first ion temperature (<span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>i</mi></mrow></msub></math></span>) measurements from the MAST-U divertor using a Retarding Field Energy Analyzer (RFEA). Embedded within the flat tile of the closed divertor chamber, the RFEA captures <span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>i</mi></mrow></msub></math></span> profiles across various plasma scenarios, including transitions to the Super-X configuration. Measurements were conducted under steady-state and transient plasma conditions characterized by a plasma current (<span><math><msub><mrow><mi>I</mi></mrow><mrow><mi>p</mi></mrow></msub></math></span>) of 750 kA, electron density (<span><math><msub><mrow><mi>n</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span>) between <span><math><mrow><mn>2</mn><mo>.</mo><mn>2</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>19</mn></mrow></msup></mrow></math></span> and <span><math><mrow><mn>4</mn><mo>.</mo><mn>4</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>19</mn></mrow></msup><mspace></mspace><msup><mrow><mtext>m</mtext></mrow><mrow><mo>−</mo><mn>3</mn></mrow></msup></mrow></math></span>, and Neutral Beam Injection (NBI) power ranging from 3.0 MW to 3.2 MW. The ion temperatures, peaking at approximately 17 eV in steady state, were compared with electron temperatures (<span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span>) obtained from Langmuir probes (LP) at identical radial positions. Preliminary findings reveal a <span><math><mrow><msub><mrow><mi>T</mi></mrow><mrow><mi>i</mi></mrow></msub><mo>/</mo><msub><mrow><mi>T</mi></mrow><mrow><mi>e</mi></mrow></msub></mrow></math></span> ratio ranging from 1 to 2.2. Additionally, high temporal resolution measurements (<span><math><mrow><mn>100</mn><mspace></mspace><mi>μ</mi><mi>s</mi></mrow></math></span>) captured the dynamics of Edge Localized Modes (ELMs), showing <span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>i</mi></mrow></msub></math></span> peaks at 16 ±1.8 eV during ELM events, nearly three times higher than inter-ELM temperatures.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"43 ","pages":"Article 101912"},"PeriodicalIF":2.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143600969","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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