Nuclear Materials and Energy最新文献

Investigation of the deposition behaviour of cobalt on 304 stainless steel in a simulated spent nuclear fuel pool 模拟乏燃料池中304不锈钢表面钴沉积行为的研究

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-27 DOI: 10.1016/j.nme.2026.102073

Jian Deng , Lin Zhong , Guolong Wang , Zeyong Lei , Mu Zhao , Jieheng Lei

The accumulation of radioactive corrosion products, specifically ⁵⁸Co and ⁶⁰Co, on metallic material (304 stainless steel) surface in spent nuclear fuel (SNF) pools is among the main factors of radioactive contamination. In this study, the microstructural characteristics and chemical composition of the surface layer of 304 stainless steel (304SS) exposed to 333 K Co-containing boric acid solution for 10, 30, 50, 70, 90, and 125 days were investigated. The cobalt deposition behaviour was analysed via material characterization techniques, E–pH diagrams, and Gibbs free energy calculations. The results revealed that CoFe₂O₄ and CoCr₂O₄ were deposited on the 304SS surface when the solution pH value was less than 6.08, and Co(OH)₂ and Co(Fe, Cr)₂O₄ were deposited on the 304SS surface when the solution pH was greater than 6.08. After 125 days of soaking, 166 nm thick Co(OH)₂ layer was deposited on the surface of 304SS, and 6 nm thick Co(Fe, Cr)₂O₄ layer beneath it. It was further analyzed that Co(OH)₂ was primarily produced by the precipitation of Co²⁺ with OH^– in solution, whereas CoFe₂O₄ and CoCr₂O₄ were primarily produced by the coprecipitation of Co²⁺ in the solution with Fe³⁺ and Cr³⁺ dissolved from the substrate. This study provides key insights into the formation mechanisms of cobalt deposition layers on 304SS in SNF pool and provides a theoretical reference for optimizing primary water chemistry, improving structural materials, and selecting decontamination strategies during operation or decommissioning.

乏燃料池中金属材料（304不锈钢）表面的放射性腐蚀产物，特别是58Co和60Co的积累是放射性污染的主要因素之一。本研究研究了304不锈钢（304SS）在333 K含钴硼酸溶液中暴露10、30、50、70、90和125天的表层显微组织特征和化学成分。通过材料表征技术、E-pH图和吉布斯自由能计算分析钴沉积行为。结果表明，当溶液pH值小于6.08时，在304SS表面沉积了CoFe2O4和CoCr2O4；当溶液pH值大于6.08时，在304SS表面沉积了Co(OH)2和Co(Fe, Cr)2O4。浸泡125 d后，304SS表面沉积了166 nm厚的Co(OH)2层，其下沉积了6 nm厚的Co(Fe, Cr)2O4层。进一步分析，Co(OH)2主要由溶液中Co2+与OH -析出产生，而CoFe2O4和CoCr2O4主要由溶液中Co2+与底物中溶解的Fe3+和Cr3+共析出产生。该研究为SNF池304SS上钴沉积层的形成机制提供了关键见解，为运行或退役过程中优化初级水化学、改进结构材料、选择去污策略提供了理论参考。

{"title":"Investigation of the deposition behaviour of cobalt on 304 stainless steel in a simulated spent nuclear fuel pool","authors":"Jian Deng , Lin Zhong , Guolong Wang , Zeyong Lei , Mu Zhao , Jieheng Lei","doi":"10.1016/j.nme.2026.102073","DOIUrl":"10.1016/j.nme.2026.102073","url":null,"abstract":"<div><div>The accumulation of radioactive corrosion products, specifically <sup>58</sup>Co and <sup>60</sup>Co, on metallic material (304 stainless steel) surface in spent nuclear fuel (SNF) pools is among the main factors of radioactive contamination. In this study, the microstructural characteristics and chemical composition of the surface layer of 304 stainless steel (304SS) exposed to 333 K Co-containing boric acid solution for 10, 30, 50, 70, 90, and 125 days were investigated. The cobalt deposition behaviour was analysed via material characterization techniques, E–pH diagrams, and Gibbs free energy calculations. The results revealed that CoFe<sub>2</sub>O<sub>4</sub> and CoCr<sub>2</sub>O<sub>4</sub> were deposited on the 304SS surface when the solution pH value was less than 6.08, and Co(OH)<sub>2</sub> and Co(Fe, Cr)<sub>2</sub>O<sub>4</sub> were deposited on the 304SS surface when the solution pH was greater than 6.08. After 125 days of soaking, 166 nm thick Co(OH)<sub>2</sub> layer was deposited on the surface of 304SS, and 6 nm thick Co(Fe, Cr)<sub>2</sub>O<sub>4</sub> layer beneath it. It was further analyzed that Co(OH)<sub>2</sub> was primarily produced by the precipitation of Co<sup>2+</sup> with OH<sup>–</sup> in solution, whereas CoFe<sub>2</sub>O<sub>4</sub> and CoCr<sub>2</sub>O<sub>4</sub> were primarily produced by the coprecipitation of Co<sup>2+</sup> in the solution with Fe<sup>3+</sup> and Cr<sup>3+</sup> dissolved from the substrate. This study provides key insights into the formation mechanisms of cobalt deposition layers on 304SS in SNF pool and provides a theoretical reference for optimizing primary water chemistry, improving structural materials, and selecting decontamination strategies during operation or decommissioning.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102073"},"PeriodicalIF":2.7,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077939","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}

引用次数: 0

Model for the nanoindentation hardness-depth relationships of ion-irradiated bicrystals with grain boundary effect 考虑晶界效应的离子辐照双晶纳米压痕硬度-深度关系模型

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-26 DOI: 10.1016/j.nme.2026.102074

Kai. Liu , Ting. Liu , Liang Xia

This study investigates the hardening mechanism of bicrystalline materials under ion irradiation by integrating experimental measurements with theoretical analysis. The experimental process involved He⁺ irradiation and nanoindentation, which revealed stage-dependent hardening behaviors. To explain the relevant experimental phenomena, a mechanistic model was developed to characterize the depth-dependent hardness evolution in ion-irradiated bicrystals. The three dominant hardening mechanisms throughout the entire nanoindentation process have been systematically analyzed for the first time, including the indentation size effect (ISE) caused by geometrically necessary dislocations (GNDs), irradiation hardening determined by inhomogeneously distributed irradiation defects, and the contribution of statistically stored dislocations (SSDs). The former two factors, influenced by the average density of dislocations and irradiation defect within the plastic zone, are affected by the s boundary (GB) and indentation depth. Considering the influence of GB on both the geometrical configuration and expansion capacity of the plastic zone, the relationship between hardness and indentation depth in ion-irradiated bicrystals was explicitly derived across four different stages. Based on this model, the evolution of associated microstructures can be quantitatively assessed, encompassing the plastic zone volume, the average density of irradiation defect and GNDs. The validity and accuracy of the proposed model were validated by comparing the theoretical results with the experimental data obtained from nanoindentation tests on double-layer copper samples. Furthermore, the model demonstrates predictive capability, with predictions showing strong agreement with experimental results.

本文采用实验测量与理论分析相结合的方法研究了离子辐照下双晶材料的硬化机理。实验过程包括He+辐照和纳米压痕，揭示了不同阶段的硬化行为。为了解释相关的实验现象，建立了一个机制模型来描述离子辐照双晶中随深度变化的硬度演变。首次系统分析了整个纳米压痕过程中的三种主要硬化机制，包括几何必要位错（GNDs）引起的压痕尺寸效应（ISE）、非均匀分布的辐照缺陷决定的辐照硬化以及统计存储位错（ssd）的贡献。前两个因素受位错平均密度和塑性区辐照缺陷的影响，受s边界（GB）和压痕深度的影响。考虑GB对塑性区的几何形态和扩展能力的影响，明确推导了离子辐照双晶中硬度与压痕深度之间的关系。基于该模型，可以定量评估相关微观组织的演变，包括塑性区体积、辐照缺陷的平均密度和GNDs。将理论结果与双层铜样品纳米压痕实验数据进行对比，验证了该模型的有效性和准确性。此外，该模型显示了预测能力，预测结果与实验结果非常吻合。

{"title":"Model for the nanoindentation hardness-depth relationships of ion-irradiated bicrystals with grain boundary effect","authors":"Kai. Liu , Ting. Liu , Liang Xia","doi":"10.1016/j.nme.2026.102074","DOIUrl":"10.1016/j.nme.2026.102074","url":null,"abstract":"<div><div>This study investigates the hardening mechanism of bicrystalline materials under ion irradiation by integrating experimental measurements with theoretical analysis. The experimental process involved He<sup>+</sup> irradiation and nanoindentation, which revealed stage-dependent hardening behaviors. To explain the relevant experimental phenomena, a mechanistic model was developed to characterize the depth-dependent hardness evolution in ion-irradiated bicrystals. The three dominant hardening mechanisms throughout the entire nanoindentation process have been systematically analyzed for the first time, including the indentation size effect (ISE) caused by geometrically necessary dislocations (GNDs), irradiation hardening determined by inhomogeneously distributed irradiation defects, and the contribution of statistically stored dislocations (SSDs). The former two factors, influenced by the average density of dislocations and irradiation defect within the plastic zone, are affected by the s boundary (GB) and indentation depth. Considering the influence of GB on both the geometrical configuration and expansion capacity of the plastic zone, the relationship between hardness and indentation depth in ion-irradiated bicrystals was explicitly derived across four different stages. Based on this model, the evolution of associated microstructures can be quantitatively assessed, encompassing the plastic zone volume, the average density of irradiation defect and GNDs. The validity and accuracy of the proposed model were validated by comparing the theoretical results with the experimental data obtained from nanoindentation tests on double-layer copper samples. Furthermore, the model demonstrates predictive capability, with predictions showing strong agreement with experimental results.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102074"},"PeriodicalIF":2.7,"publicationDate":"2026-01-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077941","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}

引用次数: 0

Comparison of D retention for advanced plasma facing materials by D ion implantation D离子注入对高级等离子体表面材料D保留的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-25 DOI: 10.1016/j.nme.2026.102069

Shingo Okumura , Yuzuka Hoshino , Ayumu Hayakawa , Kenshiro Miura , Fei Sun , Suguru Masuzaki , Makoto Oyaizu , Robert Kolasinski , Chase N. Taylor , Teppei Otsuka , Yuji Hatano , Masashi Shimada , Hao Yu , Ryuta Kasada , Akira Hasegawa , Yasuhisa Oya

For the evaluation of hydrogen isotope retention behavior for advanced plasma facing materials like W-Ta, W-Mo alloys and K-doped W, D₂⁺ implantation with different incident energies of 1 keV and 3 keV was performed up to the fluence of 1x10²² D m⁻². Thereafter D retention behavior was evaluated by thermal desorption spectroscopy (TDS) up to the temperature of 1173 K. 6 MeV Fe²⁺ irradiation was also performed to introduce the irradiation damage up to the damage level of 1 dpa, followed by the evaluation of D retention. In addition, positron annihilation spectroscopy (PAS) was performed to clarify the density and size of irradiation defects among these advanced W materials. The HIDT (Hydrogen Isotopes Diffusion and Trapping) simulation was applied to evaluate the activation energies of D trapping and their trap densities based exclusively on D₂ desorption.

The results showed that no large D retention enhancement was found for W alloys, but the D trap density with higher trap energy was reduced. In especially, the formation of large voids was refrained and D trapping by small trap energy like mono-vacancy was the major D trapping sites for K-doped W. For W-Mo and W-Ta, the addition of minor element would occupy the irradiation defects leading to the refrain of D trapping with stable D trap energy.

为了评估W- ta、W- mo合金和k掺杂W等先进等离子体表面材料的氢同位素保留行为，在1 keV和3 keV的入射能量下进行了D2+注入，注入量为1x1022 D m−2。然后用热解吸光谱（TDS）评价了D在1173 k温度下的保留行为，并进行了6 MeV Fe2+辐照，引入了1 dpa的辐照损伤水平，然后进行了D保留评价。此外，利用正电子湮没光谱（PAS）分析了这些先进W材料的辐照缺陷密度和尺寸。采用氢同位素扩散和捕获（HIDT）模拟方法，对D2脱附过程中D捕获的活化能及其捕获密度进行了计算。结果表明：W合金的D保留没有明显的增强，但陷阱能量较高的D陷阱密度降低；特别是抑制了大空洞的形成，单空位等小阱能捕获D是k掺杂w的主要捕获位点。对于W-Mo和W-Ta，少量元素的加入会占据辐照缺陷，导致以稳定的D阱能捕获D。

{"title":"Comparison of D retention for advanced plasma facing materials by D ion implantation","authors":"Shingo Okumura , Yuzuka Hoshino , Ayumu Hayakawa , Kenshiro Miura , Fei Sun , Suguru Masuzaki , Makoto Oyaizu , Robert Kolasinski , Chase N. Taylor , Teppei Otsuka , Yuji Hatano , Masashi Shimada , Hao Yu , Ryuta Kasada , Akira Hasegawa , Yasuhisa Oya","doi":"10.1016/j.nme.2026.102069","DOIUrl":"10.1016/j.nme.2026.102069","url":null,"abstract":"<div><div>For the evaluation of hydrogen isotope retention behavior for advanced plasma facing materials like W-Ta, W-Mo alloys and K-doped W, D<sub>2</sub><sup>+</sup> implantation with different incident energies of 1 keV and 3 keV was performed up to the fluence of 1x10<sup>22</sup> D m<sup>−2</sup>. Thereafter D retention behavior was evaluated by thermal desorption spectroscopy (TDS) up to the temperature of 1173 K. 6 MeV Fe<sup>2+</sup> irradiation was also performed to introduce the irradiation damage up to the damage level of 1 dpa, followed by the evaluation of D retention. In addition, positron annihilation spectroscopy (PAS) was performed to clarify the density and size of irradiation defects among these advanced W materials. The HIDT (Hydrogen Isotopes Diffusion and Trapping) simulation was applied to evaluate the activation energies of D trapping and their trap densities based exclusively on D<sub>2</sub> desorption.</div><div>The results showed that no large D retention enhancement was found for W alloys, but the D trap density with higher trap energy was reduced. In especially, the formation of large voids was refrained and D trapping by small trap energy like mono-vacancy was the major D trapping sites for K-doped W. For W-Mo and W-Ta, the addition of minor element would occupy the irradiation defects leading to the refrain of D trapping with stable D trap energy.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102069"},"PeriodicalIF":2.7,"publicationDate":"2026-01-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078031","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}

引用次数: 0

Innovative laser-based methods for monitoring fuel retention in ITER 在ITER中监测燃料保留的创新激光方法

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.nme.2026.102066

A. Huber , Ph. Andrew , G. Sergienko , J. Assmann , D. Castano , A. De Schepper , S. Friese , R. Greven , I. Ivashov , D. Kampf , Y. Krasikov , C.C. Klepper , H.T. Lambertz , Ph. Mertens , K. Mlynczak , G. Offermanns , B. Quinlan , K. Rasinska , M. Schrader , D. Van Staden , Ch. Linsmeier

This paper addresses the challenge of tritium inventory management in ITER and future fusion reactors, highlighting the importance of accurate tritium measurement and its spatial distribution within the vacuum vessel. Given ITER’s operational constraints, especially the limit on tritium retention, precise measurement is essential for both safety and regulatory compliance. To tackle these questions, the paper presents the T-monitor diagnostic system developed by Forschungszentrum Jülich, which uses Laser-Induced Desorption (LID) in combination with Diagnostic Residual Gas Analysis (DRGA) to measure hydrogen isotope concentrations on the surface of divertor tiles. The system integrates a high-power laser, advanced optical components, and a Fast Scanning Mirror Unit (FSMU) for accurate laser spot positioning with rapid response.

Designed to measure in situ tritium retention, the diagnostic provides high-resolution spatial mapping, vital for evaluating detritiation strategies. The laser heating process increases the divertor surface temperature to 1600 K within the laser spot, promoting hydrogen isotope desorption. Accurate measurements require the precise control of laser parameters, including pulse duration and spot size, with a target relative accuracy of 20%. The optical design includes both in-vessel and ex-vessel components, such as durable high-reflectivity mirrors made of gold and copper, selected not only for their infrared performance but also for their transmission of visible wavelengths for observation purposes. To protect optical components from contamination, a pneumatic shutter is used.

本文讨论了ITER和未来聚变反应堆中氚库存管理的挑战，强调了准确测量氚及其在真空容器内空间分布的重要性。考虑到ITER的运行限制，特别是氚保留的限制，精确的测量对于安全性和法规遵从性都是至关重要的。为了解决这些问题，本文介绍了由Forschungszentrum j lich开发的T-monitor诊断系统，该系统使用激光诱导解吸（LID）结合诊断残余气体分析（DRGA）来测量导流瓦表面的氢同位素浓度。该系统集成了高功率激光器、先进的光学元件和快速扫描镜单元（FSMU），用于快速响应的精确激光光斑定位。设计用于测量原位氚保留，诊断提供高分辨率的空间映射，对评估氚化策略至关重要。激光加热过程使导流器表面温度提高到1600 K，促进了氢同位素的解吸。精确测量需要精确控制激光参数，包括脉冲持续时间和光斑大小，目标相对精度为20%。光学设计包括容器内和容器外组件，例如由黄金和铜制成的耐用高反射率镜子，不仅因其红外性能而被选中，而且还因其可见光传输而被选中用于观测目的。为了保护光学元件不受污染，使用了气动快门。

{"title":"Innovative laser-based methods for monitoring fuel retention in ITER","authors":"A. Huber , Ph. Andrew , G. Sergienko , J. Assmann , D. Castano , A. De Schepper , S. Friese , R. Greven , I. Ivashov , D. Kampf , Y. Krasikov , C.C. Klepper , H.T. Lambertz , Ph. Mertens , K. Mlynczak , G. Offermanns , B. Quinlan , K. Rasinska , M. Schrader , D. Van Staden , Ch. Linsmeier","doi":"10.1016/j.nme.2026.102066","DOIUrl":"10.1016/j.nme.2026.102066","url":null,"abstract":"<div><div>This paper addresses the challenge of tritium inventory management in ITER and future fusion reactors, highlighting the importance of accurate tritium measurement and its spatial distribution within the vacuum vessel. Given ITER’s operational constraints, especially the limit on tritium retention, precise measurement is essential for both safety and regulatory compliance. To tackle these questions, the paper presents the T-monitor diagnostic system developed by Forschungszentrum Jülich, which uses Laser-Induced Desorption (LID) in combination with Diagnostic Residual Gas Analysis (DRGA) to measure hydrogen isotope concentrations on the surface of divertor tiles. The system integrates a high-power laser, advanced optical components, and a Fast Scanning Mirror Unit (FSMU) for accurate laser spot positioning with rapid response.</div><div>Designed to measure <em>in situ</em> tritium retention, the diagnostic provides high-resolution spatial mapping, vital for evaluating detritiation strategies. The laser heating process increases the divertor surface temperature to 1600 K within the laser spot, promoting hydrogen isotope desorption. Accurate measurements require the precise control of laser parameters, including pulse duration and spot size, with a target relative accuracy of 20%. The optical design includes both in-vessel and ex-vessel components, such as durable high-reflectivity mirrors made of gold and copper, selected not only for their infrared performance but also for their transmission of visible wavelengths for observation purposes. To protect optical components from contamination, a pneumatic shutter is used.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102066"},"PeriodicalIF":2.7,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078032","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}

引用次数: 0

Quantification of deuterium and low-Z impurity deposition on long-term samples exposed in ASDEX Upgrade 长期暴露于ASDEX Upgrade的样品中氘和低z杂质沉积的定量分析

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.nme.2026.102070

K. Krieger, V. Rohde, ASDEX Upgrade Team

Long Term Sample (LTS) holders installed at remote wall locations of ASDEX Upgrade (AUG) and equipped with silicon wafer witness samples are used to monitor the integral net deposition of boron by glow discharge boronisation (GDB). Deposited low-Z impurities (B, C) and deuterium are quantified by

^{3} He

Nuclear Reaction Analysis. Isotopes are resolved by analysis of their characteristic peaks in the energy spectra of detected protons. For the analysis a fitting procedure, tailored to the expected species, has been implemented and calibrated against quantitatively characterised standard samples.

The analysis has been applied to LTS from recent AUG campaigns. The results revealed that only a small fraction of the total boron introduced during the campaign can be accounted for with the rest assumed to be exhausted as molecular compounds during the campaign and also during post-campaign flushing of the vessel with air.

长期样品（LTS）支架安装在ASDEX Upgrade （AUG）的远程壁位置，并配备硅片见证样品，用于通过辉光放电硼化（GDB）监测硼的整体净沉积。沉积的低z杂质（B， C）和氘用3He核反应分析定量。同位素是通过分析它们在探测到的质子能谱中的特征峰来分辨的。对于分析，针对预期物种量身定制的拟合程序已经实施，并针对定量表征的标准样品进行了校准。该分析已应用于近期AUG战役中的LTS。结果表明，在整个过程中引入的硼中只有一小部分可以解释，其余部分被认为是在整个过程中作为分子化合物消耗掉的，并且在活动后用空气冲洗容器时也是如此。

引用次数: 0

Review on ASDEX Upgrade operation with tungsten plasma facing components 钨等离子表面组件ASDEX升级操作回顾

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-19 DOI: 10.1016/j.nme.2026.102063

R. Neu, C. Angioni, V. Bobkov, R. Dux, J. Hobirk, A. Kallenbach, K. Krieger, T. Pütterich, V. Rohde, K. Schmid

With the decision of ITER to start its operation with tungsten as plasma facing material also for the main chamber plasma facing components, the interest in the consequences of the use of W was strongly increased. Although many investigations had already been carried out in the all-W ASDEX Upgrade, EAST and WEST tokamak experiments, the ITER decision raised many new questions. To provide a robust foundation for addressing these questions, this paper reviews the tungsten related investigations carried out in ASDEX Upgrade over the last three decades in order to summarize the results achieved so far and to provide a comprehensive list of references for more detailed reading. Further to conclude from this material which results can be used directly for the full-W ITER, where further work is needed and possibly rewarding and which areas are difficult to be researched in AUG and other present-day devices.

随着ITER决定将钨作为等离子体面材料并用于主室等离子体面组件开始运行，人们对使用钨的后果的兴趣大大增加。尽管在全w ASDEX升级、东、西托卡马克实验中已经进行了许多研究，但ITER的决定提出了许多新的问题。为了为解决这些问题提供一个坚实的基础，本文回顾了过去三十年来在ASDEX Upgrade中进行的钨相关研究，以总结迄今为止取得的成果，并提供了一份全面的参考资料清单，以供更详细的阅读。进一步从这些材料中得出结论，哪些结果可以直接用于全w ITER，哪些需要进一步的工作并且可能有回报，哪些领域在AUG和其他现有设备中难以研究。

引用次数: 0

Effects of neutron irradiation on the thermionic emission properties of LaB6-emitters used in neutral gas pressure gauges 中子辐照对中性气体压力表用lab6发射体热离子发射性能的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-19 DOI: 10.1016/j.nme.2026.102068

V. Haak , A. Bukowicka , A. Graband , K.D. Hanke , T. Hannappel , J. Koch , G. Motojima , E. Morishita , D. Naujoks , T. Stummer , T. Sturm , M. Tokitani , M. Villa

After the failure of a LaB

_{6}

-emitter used in a neutral gas pressure gauge in the Large Helical Device during deuterium operation, the effect of neutrons on the thermionic emission properties of LaB

_{6}

is studied in this work. For that purpose, six cylindrical LaB

_{6}

-samples of 8 mm length and 1 mm diameter were irradiated with neutron doses between

1 \cdot 1 0^{16} - 3 \cdot 1 0^{17}

\frac{n}{{cm}^{2}}

at the TRIGA Mark-II reactor. Analysis after neutron irradiation showed accumulation of oxygen on the sample surfaces and the formation of cracks, holes, growth of surface layers and erosion at the sample edges. In order to study the thermionic emission properties of the neutron-irradiated LaB

_{6}

-samples, they were used as emitters in a neutral gas pressure gauge under vacuum conditions and in a magnetic field. The tests revealed significantly reduced thermionic emission directly after neutron irradiation that improves when repeating the measurements. Four out of six LaB

_{6}

-emitters eventually reach thermionic emission properties comparable to the reference emitters again, due to either the removal of the lanthanum-oxide surface layer from the emitter surface during operation in the neutral gas pressure gauge or the thermal recovery of neutron-induced lattice defects.

针对大型螺旋装置中用于中性气压力表的LaB6发射器在氘运行过程中发生的故障，研究了中子对LaB6热离子发射特性的影响。为此，在TRIGA Mark-II反应堆上，以1⋅1016−3⋅1017 ncm2的中子剂量照射6个长度为8 mm、直径为1 mm的圆柱形lab6样品。中子辐照后的分析表明，样品表面氧气积累，形成裂纹、孔洞，表层生长，样品边缘有侵蚀。为了研究中子辐照后lab6样品的热离子发射特性，分别在真空条件下和磁场条件下的中性气体压力表中作为发射体。试验表明，中子辐照后直接热离子发射显著降低，重复测量后效果改善。6个lab6发射体中有4个最终达到了与参考发射体相当的热离子发射性能，这要么是由于在中性气体压力表中操作时从发射体表面去除了氧化镧表面层，要么是由于中子引起的晶格缺陷的热恢复。

{"title":"Effects of neutron irradiation on the thermionic emission properties of LaB6-emitters used in neutral gas pressure gauges","authors":"V. Haak , A. Bukowicka , A. Graband , K.D. Hanke , T. Hannappel , J. Koch , G. Motojima , E. Morishita , D. Naujoks , T. Stummer , T. Sturm , M. Tokitani , M. Villa","doi":"10.1016/j.nme.2026.102068","DOIUrl":"10.1016/j.nme.2026.102068","url":null,"abstract":"<div><div>After the failure of a LaB<span><math><msub><mrow></mrow><mrow><mn>6</mn></mrow></msub></math></span>-emitter used in a neutral gas pressure gauge in the Large Helical Device during deuterium operation, the effect of neutrons on the thermionic emission properties of LaB<span><math><msub><mrow></mrow><mrow><mn>6</mn></mrow></msub></math></span> is studied in this work. For that purpose, six cylindrical LaB<span><math><msub><mrow></mrow><mrow><mn>6</mn></mrow></msub></math></span>-samples of 8 mm length and 1 mm diameter were irradiated with neutron doses between <span><math><mrow><mn>1</mn><mi>⋅</mi><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>16</mn></mrow></msup><mo>−</mo><mn>3</mn><mi>⋅</mi><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>17</mn></mrow></msup></mrow></math></span> <!--> <span><math><mfrac><mrow><mtext>n</mtext></mrow><mrow><msup><mrow><mtext>cm</mtext></mrow><mrow><mn>2</mn></mrow></msup></mrow></mfrac></math></span> at the TRIGA Mark-II reactor. Analysis after neutron irradiation showed accumulation of oxygen on the sample surfaces and the formation of cracks, holes, growth of surface layers and erosion at the sample edges. In order to study the thermionic emission properties of the neutron-irradiated LaB<span><math><msub><mrow></mrow><mrow><mn>6</mn></mrow></msub></math></span>-samples, they were used as emitters in a neutral gas pressure gauge under vacuum conditions and in a magnetic field. The tests revealed significantly reduced thermionic emission directly after neutron irradiation that improves when repeating the measurements. Four out of six LaB<span><math><msub><mrow></mrow><mrow><mn>6</mn></mrow></msub></math></span>-emitters eventually reach thermionic emission properties comparable to the reference emitters again, due to either the removal of the lanthanum-oxide surface layer from the emitter surface during operation in the neutral gas pressure gauge or the thermal recovery of neutron-induced lattice defects.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102068"},"PeriodicalIF":2.7,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146023234","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}

引用次数: 0

High heat flux testing of actively cooled graphite- and tungsten-armoured JT-60SA flat tile divertor mock-ups 主动冷却石墨和钨铠装JT-60SA平铺瓦导流器模型的高热流密度测试

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-17 DOI: 10.1016/j.nme.2026.102067

Daniel Dickes , Bernd Böswirth , Katja Hunger , Gabriel Peyron , Quentin Tichit , Marianne Richou , Johann Riesch , Mehdi Firdaouss , Valerio Tomarchio , Rudolf Neu

During the process of developing actively cooled divertor plasma-facing components for JT-60SA, a fusion experiment in Japan built within the framework of the “Broader Approach Agreement” between the European Union and Japan, three small-scale divertor mock-ups have been manufactured. The mock-ups follow the flat-tile design, i.e., have plasma-facing armour tiles joined to an actively cooled heat sink. One mock-up has tungsten (W) armour tiles, and two mock-ups have carbon (C) armour tiles, with the heat sink material being the molybdenum alloy TZM. The joining was realized via diffusion bonding with a titanium interlayer. In this work, the thermo-mechanical behaviour of the mock-ups is assessed with the high heat flux test facility GLADIS in order to qualify the joining technology. This includes screening tests up to a heat flux of 15 MW/m² and cyclic loading with a heat flux of 10 MW/m² for 10 s. During high heat flux testing, pyrometer and thermocouple temperature measurements, digital camera images, and thermographic imaging were used to monitor the mock-ups. In addition, comparative infrared thermography tests and visual characterizations before and after high heat flux testing were performed, including the preparation of cross-sections for scanning electron microscopy. Debonding of the armour tiles did not occur during high heat flux testing, indicating that the diffusion bonding process is suitable. However, this work outlines challenges like a potentially decreasing heat removal capability of the TZM heat sink during cyclic loading or the occurrence of detrimental deep cracking in the case of W armour tiles.

在为JT-60SA开发主动冷却分流器面向等离子体组件的过程中，在欧盟和日本之间的“更广泛的方法协议”框架内建立的日本聚变实验中，已经制造了三个小型分流器模型。模型采用平瓦设计，也就是说，有等离子体面装甲瓦连接到一个主动冷却散热器。一艘实物模型采用钨(W)装甲瓦，两艘实物模型采用碳(C)装甲瓦，散热器材料为钼合金TZM。连接是通过钛中间层的扩散连接实现的。在这项工作中，用高热流密度测试设备GLADIS评估了模型的热力学行为，以确定连接技术的合格性。这包括热流密度高达15 MW/m2的筛选试验和热流密度为10 MW/m2的循环加载10 s。在高热流密度测试期间，使用高温计和热电偶温度测量，数码相机图像和热成像来监测模型。此外，还进行了对比红外热成像测试和高热流密度测试前后的视觉表征，包括扫描电子显微镜的截面制备。在高热流密度试验中，装甲瓦未发生脱粘现象，表明扩散粘接工艺是合适的。然而，这项工作概述了一些挑战，比如在循环加载期间，TZM散热器的散热能力可能会下降，或者W装甲瓦的情况下会发生有害的深度开裂。

{"title":"High heat flux testing of actively cooled graphite- and tungsten-armoured JT-60SA flat tile divertor mock-ups","authors":"Daniel Dickes , Bernd Böswirth , Katja Hunger , Gabriel Peyron , Quentin Tichit , Marianne Richou , Johann Riesch , Mehdi Firdaouss , Valerio Tomarchio , Rudolf Neu","doi":"10.1016/j.nme.2026.102067","DOIUrl":"10.1016/j.nme.2026.102067","url":null,"abstract":"<div><div>During the process of developing actively cooled divertor plasma-facing components for JT-60SA, a fusion experiment in Japan built within the framework of the “Broader Approach Agreement” between the European Union and Japan, three small-scale divertor mock-ups have been manufactured. The mock-ups follow the flat-tile design, i.e., have plasma-facing armour tiles joined to an actively cooled heat sink. One mock-up has tungsten (W) armour tiles, and two mock-ups have carbon (C) armour tiles, with the heat sink material being the molybdenum alloy TZM. The joining was realized via diffusion bonding with a titanium interlayer. In this work, the thermo-mechanical behaviour of the mock-ups is assessed with the high heat flux test facility GLADIS in order to qualify the joining technology. This includes screening tests up to a heat flux of 15 MW/m<sup>2</sup> and cyclic loading with a heat flux of 10 MW/m<sup>2</sup> for 10 s. During high heat flux testing, pyrometer and thermocouple temperature measurements, digital camera images, and thermographic imaging were used to monitor the mock-ups. In addition, comparative infrared thermography tests and visual characterizations before and after high heat flux testing were performed, including the preparation of cross-sections for scanning electron microscopy. Debonding of the armour tiles did not occur during high heat flux testing, indicating that the diffusion bonding process is suitable. However, this work outlines challenges like a potentially decreasing heat removal capability of the TZM heat sink during cyclic loading or the occurrence of detrimental deep cracking in the case of W armour tiles.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102067"},"PeriodicalIF":2.7,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146023235","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}

引用次数: 0

Modification of chemical and mechanical properties of p-W-O coating after Magnum-PSI D2-N2 plasma exposure and its consequences for the analysis of LIBS spectra Magnum-PSI D2-N2等离子体辐照后p-W-O涂层化学力学性能的改变及其对LIBS光谱分析的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-17 DOI: 10.1016/j.nme.2026.102065

Indrek Jõgi , Peeter Paris , Kaarel Piip , Jordy Vernimmen , Beata Tyburska-Pueschel , Sven Lange , Taivo Jõgiaas , Matteo Passoni , David Dellasega , Gabriele Alberti , Hennie van der Meiden

The present study investigated the effect of D₂-N₂ (7%) plasma exposure in Magnum-PSI on the D retention and chemical and mechanical properties of a porous W-O (p-W-O) coating. The variation of the chemical composition, crystalline phase and mechanical properties along the sample surface were determined by Nuclear Reaction Analysis (NRA), Rutherford Backscattering Spectroscopy (RBS), nanoindentation and Raman spectroscopy. These changes were compared to the Laser-Induced Breakdown Spectroscopy (LIBS) measurements. LIBS depth profiles of W and Mo were consistent with the profiles determined by NRA and RBS, showing a W-O layer, a thin W adhesion layer and a Mo substrate. Typically, the high D intensity was determined only during the first LIBS laser shot on a measurement spot, while the spatial distribution of D intensity determined by LIBS along the coating surface followed the D concentration determined by NRA. According to the Raman spectra, the investigated p-W-O coating corresponded to nanograins of W-O and the phase composition was relatively uniform along the coating surface. The elastic modulus of p-W-O coating was considerably lower than the modulus of Mo coating or bulk W coating and corresponded to the values found in other studies carried out with W-O mixtures. The elastic modulus of p-W-O coating decreased towards the edge of the coating. The study revealed that the modulus and the background intensity of the LIBS spectra had a negative correlation, suggesting that LIBS may be a suitable method for the estimation of the stiffness of tungsten co-deposits as a similar correlation is shown for other types of W coatings.

本研究研究了在Magnum-PSI中暴露D2-N2（7%）等离子体对多孔W-O （p-W-O）涂层的D保留和化学力学性能的影响。采用核反应分析（NRA）、卢瑟福后向散射光谱（RBS）、纳米压痕和拉曼光谱测定了样品表面化学成分、晶相和力学性能的变化。这些变化与激光诱导击穿光谱（LIBS）测量结果进行了比较。W和Mo的LIBS深度分布图与NRA和RBS测定的分布图一致，呈现W- o层、薄W附着层和Mo衬底。通常情况下，高D强度仅在第一次LIBS激光射入测点时产生，而LIBS测定的D强度沿涂层表面的空间分布遵循NRA测定的D浓度。拉曼光谱表明，所制备的p-W-O涂层与W-O纳米颗粒相对应，涂层表面相组成相对均匀。p-W-O涂层的弹性模量大大低于Mo涂层或块状W涂层的弹性模量，与W- o混合物的其他研究结果一致。p-W-O涂层的弹性模量沿涂层边缘逐渐减小。研究表明，LIBS光谱的模量与背景强度呈负相关，表明LIBS可能是估计钨共镀层硬度的合适方法，因为其他类型的W涂层也显示出类似的相关性。

{"title":"Modification of chemical and mechanical properties of p-W-O coating after Magnum-PSI D2-N2 plasma exposure and its consequences for the analysis of LIBS spectra","authors":"Indrek Jõgi , Peeter Paris , Kaarel Piip , Jordy Vernimmen , Beata Tyburska-Pueschel , Sven Lange , Taivo Jõgiaas , Matteo Passoni , David Dellasega , Gabriele Alberti , Hennie van der Meiden","doi":"10.1016/j.nme.2026.102065","DOIUrl":"10.1016/j.nme.2026.102065","url":null,"abstract":"<div><div>The present study investigated the effect of D<sub>2</sub>-N<sub>2</sub> (7%) plasma exposure in Magnum-PSI on the D retention and chemical and mechanical properties of a porous W-O (p-W-O) coating. The variation of the chemical composition, crystalline phase and mechanical properties along the sample surface were determined by Nuclear Reaction Analysis (NRA), Rutherford Backscattering Spectroscopy (RBS), nanoindentation and Raman spectroscopy. These changes were compared to the Laser-Induced Breakdown Spectroscopy (LIBS) measurements. LIBS depth profiles of W and Mo were consistent with the profiles determined by NRA and RBS, showing a W-O layer, a thin W adhesion layer and a Mo substrate. Typically, the high D intensity was determined only during the first LIBS laser shot on a measurement spot, while the spatial distribution of D intensity determined by LIBS along the coating surface followed the D concentration determined by NRA. According to the Raman spectra, the investigated p-W-O coating corresponded to nanograins of W-O and the phase composition was relatively uniform along the coating surface. The elastic modulus of p-W-O coating was considerably lower than the modulus of Mo coating or bulk W coating and corresponded to the values found in other studies carried out with W-O mixtures. The elastic modulus of p-W-O coating decreased towards the edge of the coating. The study revealed that the modulus and the background intensity of the LIBS spectra had a negative correlation, suggesting that LIBS may be a suitable method for the estimation of the stiffness of tungsten co-deposits as a similar correlation is shown for other types of W coatings.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102065"},"PeriodicalIF":2.7,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146022935","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}

引用次数: 0

Overview of advanced plasma-facing materials testing for Fusion Pilot Plants at DIII-D DIII-D聚变中试工厂先进等离子体表面材料测试概述

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-01-14 DOI: 10.1016/j.nme.2026.102064

Jonathan Coburn , Florian Effenberg , Mary Alice Cusentino , Chase Hargrove , Mykola Ialovega , Maria Morbey , Lauren Nuckols , Žana Popović , Zachary Bergstrom , Shawn Zamperini , Tyler Abrams , Dmitry Rudakov , Shota Abe , Shane Evans , Tatsuya Hinoki , Ryan Hood , Eric Lang , Charlie Lasnier , Ulises Losada , Claudio Marini , Weicheng Zhong

Characterization and testing of advanced plasma-facing materials (PFMs) for Fusion Pilot Plants (FPP) is being conducted at the DIII-D National Fusion Facility through the ongoing two-year FPP Candidate Materials Thrust. Year one tested 17 novel materials utilizing the Divertor Materials Evaluation System (DiMES), with samples analyzed pre- and post-experiment via SEM, EDS, and confocal microscopy. Repeatable reference discharges were developed to ensure uniformity between experiments, including a new strike-point rastering scenario to provide more uniform heat/particle flux across DiMES during ELMing H-mode discharges. Various sample geometries and temperatures were used to achieve FPP-relevant conditions, including samples angled 10° towards the incident plasma flux and pre-heating up to 500 °C.

The first exposure of liquid lithium (Li) capillary porous structures in a tokamak demonstrated uniform emission of Li vapor and suppression of Li droplets in H-mode when preheated to 350 °C. Dispersoid-strengthened W with 1 wt% TaC, TiC, and ZrC exposed to H-mode showed cracking and dispersoid ejection for all varieties except TiC, providing a clear down-selection. Ultra-high temperature ceramic materials TiB₂ and ZrB₂ showed minimal degradation under L-mode exposure. Silicon carbide (SiC) fiber composites showed arcing along edges, while CVD SiC remained pristine. Atmospheric plasma-sprayed W and SiC coatings endured H-mode exposure without macroscopic delamination; SiC exhibited granular ejection, while W showed increased outgassing. Additional W-based alloys were stress tested in H-mode, including Ni-based W heavy alloys, W_fSiC_f/W composites, W multi-principle element alloys, and functionally-graded W/SiC, to varying degrees of success.

DIII-D国家聚变设施正在通过正在进行的为期两年的FPP候选材料推力，对聚变中试工厂（FPP）的先进等离子体表面材料（pfm）进行表征和测试。第一年使用DiMES （Divertor materials Evaluation System）测试了17种新材料，并通过扫描电镜（SEM）、能谱仪（EDS）和共聚焦显微镜对实验前后的样品进行了分析。开发了可重复的参考放电以确保实验之间的均匀性，包括一个新的击点光栅场景，以在ELMing h模式放电期间提供更均匀的热/粒子通量。不同的样品几何形状和温度被用来达到fpp相关的条件，包括样品与入射等离子体通量成10°角，预热到500°C。首次在托卡马克中暴露液态锂（Li）毛细孔结构，当预热到350℃时，在h模式下Li蒸气均匀发射，Li液滴被抑制。添加1 wt% TaC、TiC和ZrC的弥散增强W在h模式下，除TiC外，所有品种均出现开裂和弥散弹射，提供了明确的向下选择。超高温陶瓷材料TiB2和ZrB2在l模式下的降解最小。碳化硅（SiC）纤维复合材料的边缘呈弧形，而CVD SiC则保持原始状态。大气等离子喷涂W和SiC涂层可承受h模式暴露而无宏观分层；SiC表现为颗粒状喷射，W表现为脱气增加。在h模式下对其他W基合金进行了应力测试，包括ni基W重合金、WfSiCf/W复合材料、W多元素合金和功能梯度W/SiC，均取得了不同程度的成功。

{"title":"Overview of advanced plasma-facing materials testing for Fusion Pilot Plants at DIII-D","authors":"Jonathan Coburn , Florian Effenberg , Mary Alice Cusentino , Chase Hargrove , Mykola Ialovega , Maria Morbey , Lauren Nuckols , Žana Popović , Zachary Bergstrom , Shawn Zamperini , Tyler Abrams , Dmitry Rudakov , Shota Abe , Shane Evans , Tatsuya Hinoki , Ryan Hood , Eric Lang , Charlie Lasnier , Ulises Losada , Claudio Marini , Weicheng Zhong","doi":"10.1016/j.nme.2026.102064","DOIUrl":"10.1016/j.nme.2026.102064","url":null,"abstract":"<div><div>Characterization and testing of advanced plasma-facing materials (PFMs) for Fusion Pilot Plants (FPP) is being conducted at the DIII-D National Fusion Facility through the ongoing two-year FPP Candidate Materials Thrust. Year one tested 17 novel materials utilizing the Divertor Materials Evaluation System (DiMES), with samples analyzed pre- and post-experiment via SEM, EDS, and confocal microscopy. Repeatable reference discharges were developed to ensure uniformity between experiments, including a new strike-point rastering scenario to provide more uniform heat/particle flux across DiMES during ELMing H-mode discharges. Various sample geometries and temperatures were used to achieve FPP-relevant conditions, including samples angled 10° towards the incident plasma flux and pre-heating up to 500°C.</div><div>The first exposure of liquid lithium (Li) capillary porous structures in a tokamak demonstrated uniform emission of Li vapor and suppression of Li droplets in H-mode when preheated to 350 °C. Dispersoid-strengthened W with 1 wt% TaC, TiC, and ZrC exposed to H-mode showed cracking and dispersoid ejection for all varieties except TiC, providing a clear down-selection. Ultra-high temperature ceramic materials TiB<sub>2</sub> and ZrB<sub>2</sub> showed minimal degradation under L-mode exposure. Silicon carbide (SiC) fiber composites showed arcing along edges, while CVD SiC remained pristine. Atmospheric plasma-sprayed W and SiC coatings endured H-mode exposure without macroscopic delamination; SiC exhibited granular ejection, while W showed increased outgassing. Additional W-based alloys were stress tested in H-mode, including Ni-based W heavy alloys, W<sub>f</sub>SiC<sub>f</sub>/W composites, W multi-principle element alloys, and functionally-graded W/SiC, to varying degrees of success.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"46 ","pages":"Article 102064"},"PeriodicalIF":2.7,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146077940","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}

引用次数: 0