Pub Date : 2025-01-23DOI: 10.1016/j.nme.2025.101882
Nanyu Mou , Qianqian Lin , Mingchi Feng , Shuai Huang , Le Han , Damao Yao
Due to the insolubility between W and Cu and the inability to generate intermetallic compounds, connecting W and Cu poses a considerable challenge. In this study, a casting assisted vacuum hot pressing (VHP) method is adopted to join W and Cu, and the process parameters are optimized. The casting temperature is 1180 °C and the most suitable bonding parameters are a bonding temperature of 650 °C, a bonding pressure of 25 MPa and a bonding time of 90 min. The obtained W/Cu joint exhibits an average shear strength of 138.0 MPa and with no detectable defects on the interface. The effective bonding area of W/Cu joints exceed 98 %. The thermal conductivity of W/Cu prepared by casting assisted VHP is 347, 288, and 257 W/m·K at room temperature, 500 °C, and 900 °C, respectively, exceeding that of W/Cu prepared by casting assisted hot isostatic pressing (HIP) at the corresponding temperatures. The heat transfer performance and structural integrity of the prepared divertor mockup remain satisfactory after 1000 cycles at 20 MW/m2, with a peak surface temperature of 725.7 ℃. The proposed method provides a new strategy for preparing high-performance and fatigue resistant W/Cu joints, to address the issues of low bonding performance as well as poor fatigue resistance.
{"title":"A new method for preparing high-quality and fatigue-resistant divertor W/Cu joints","authors":"Nanyu Mou , Qianqian Lin , Mingchi Feng , Shuai Huang , Le Han , Damao Yao","doi":"10.1016/j.nme.2025.101882","DOIUrl":"10.1016/j.nme.2025.101882","url":null,"abstract":"<div><div>Due to the insolubility between W and Cu and the inability to generate intermetallic compounds, connecting W and Cu poses a considerable challenge. In this study, a casting assisted vacuum hot pressing (VHP) method is adopted to join W and Cu, and the process parameters are optimized. The casting temperature is 1180 °C and the most suitable bonding parameters are a bonding temperature of 650 °C, a bonding pressure of 25 MPa and a bonding time of 90 min. The obtained W/Cu joint exhibits an average shear strength of 138.0 MPa and with no detectable defects on the interface. The effective bonding area of W/Cu joints exceed 98 %. The thermal conductivity of W/Cu prepared by casting assisted VHP is 347, 288, and 257 W/m·K at room temperature, 500 °C, and 900 °C, respectively, exceeding that of W/Cu prepared by casting assisted hot isostatic pressing (HIP) at the corresponding temperatures. The heat transfer performance and structural integrity of the prepared divertor mockup remain satisfactory after 1000 cycles at 20 MW/m<sup>2</sup>, with a peak surface temperature of 725.7 ℃. The proposed method provides a new strategy for preparing high-performance and fatigue resistant W/Cu joints, to address the issues of low bonding performance as well as poor fatigue resistance.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101882"},"PeriodicalIF":2.3,"publicationDate":"2025-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173568","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 : 2025-01-20DOI: 10.1016/j.nme.2025.101878
Hui Wang , Guoliang Xu , Rui Ding , Hang Si , Guozhang Jia , Jin Guo , Jinheng Zhao , Junling Chen
Neon (Ne) and argon (Ar) impurity seedings are recognized as effective techniques for reducing the divertor heat load in high-power operations of future reactors like CFETR. However, their influence on tungsten (W) erosion and edge transport remains unclear. Based on previously published SOLPS-ITER simulation results ([H. Si et al 2022 Nucl. Fusion 62 026031]), this study investigates the effects of Ne and Ar seedings on the W source and edge transport under partially detached divertor conditions on CFETR by DIVIMP simulations. The transport processes of Ne and Ar impurities in the boundary plasma are evaluated, with Ar expected to have better compression. Under similar partial detachment divertor conditions with the same total radiation power loss, the W leakage ability from divertor to the core plasma is demonstrated to be similar for Ne and Ar seeding cases. However, the W density in the core plasma is approximately four times lower in the Ar seeding case compared to the Ne seeding case, primarily due to the smaller W source from the divertor target. Therefore, the Ar seeding is proved to be more beneficial for reducing the W source and core W concentration than the Ne seeding.
{"title":"Comparative studies of tungsten erosion and edge transport under partially detached divertor condition of CFETR with neon and argon seedings","authors":"Hui Wang , Guoliang Xu , Rui Ding , Hang Si , Guozhang Jia , Jin Guo , Jinheng Zhao , Junling Chen","doi":"10.1016/j.nme.2025.101878","DOIUrl":"10.1016/j.nme.2025.101878","url":null,"abstract":"<div><div>Neon (Ne) and argon (Ar) impurity seedings are recognized as effective techniques for reducing the divertor heat load in high-power operations of future reactors like CFETR. However, their influence on tungsten (W) erosion and edge transport remains unclear. Based on previously published SOLPS-ITER simulation results ([H. Si et al 2022 Nucl. Fusion 62 026031]), this study investigates the effects of Ne and Ar seedings on the W source and edge transport under partially detached divertor conditions on CFETR by DIVIMP simulations. The transport processes of Ne and Ar impurities in the boundary plasma are evaluated, with Ar expected to have better compression. Under similar partial detachment divertor conditions with the same total radiation power loss, the W leakage ability from divertor to the core plasma is demonstrated to be similar for Ne and Ar seeding cases. However, the W density in the core plasma is approximately four times lower in the Ar seeding case compared to the Ne seeding case, primarily due to the smaller W source from the divertor target. Therefore, the Ar seeding is proved to be more beneficial for reducing the W source and core W concentration than the Ne seeding.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101878"},"PeriodicalIF":2.3,"publicationDate":"2025-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173099","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 : 2025-01-19DOI: 10.1016/j.nme.2025.101877
P. Dinca , C. Staicu , B. Butoi , B.G. Solomonea , A. Anghel , O.G. Pompilian , G. Bulai , V. Tiron
<div><div>The purpose of this work is to carry out a systematic parametric study on neon (Ne) retention and release behaviour from Be/Ne layers co-deposited under ITER-relevant conditions, in terms of deposition temperature, gas pressure and ion energy. This study incorporates analyses of structural and morphological properties of Be/Ne coatings, along with Ne retention/release behaviour. A total of 27 batches of Be/Ne layers were deposited onto silicon (Si) and tungsten (W) substrates by adjusting to several selected values (<em>i</em>) the working gas pressure (1 Pa, 2 Pa, 3 Pa), (<em>ii</em>) the ion energy of the plasma species (30 eV, 100 eV, 200 eV) and (<em>iii</em>) the substrate temperature during deposition (340 K, 473 K, 573 K). The chosen deposition method was the novel Bipolar-High Power Impulse Magnetron Sputtering (BP-HiPIMS) technique due to its high ionization degree of the plasma species and its versatility in controlling the ion flux and energy during the deposition process. For a more comprehensive understanding, of how plasma properties, coatings’ structure, retention and release behaviour of Ne from the deposited samples correlate to each other, plasma-diagnosis was performed. The influence of gas pressure and pulsing configuration on ion energy and flux was investigated. The coating’s morphology and microstructure were analysed by Scanning Electron Microscopy (SEM), Atomic Force Microscopy (AFM) and X-ray diffraction (XRD). The surface’s morphology of the Be-Ne layers deposited at 340 K indicates smooth layers, with surface roughness values independent of the working gas pressure and ion energy. SEM images of the layers produced at temperatures higher than 340 K suggest the formation of blister-like structures on their surface. Unsurprisingly, the crystalline structure is strongly dependent on the substrate temperature. Metallic Be crystalline phase is observed for layers deposited at 473 K and 573 K, while, for the lowest substrate temperature (340 K), the structure of the Be-Ne layers was found to be amorphous, regardless of the working gas pressure and/or ion energy used during the deposition process. Ne/Be ratio in the deposited layers was evaluated through Thermal Desorption Spectroscopy (TDS) measurements. The results show an increase in the Ne content in layers with the increase of working gas pressure and ion energy. Although the Ne inventory is released above 1000 K, the increase of substrate temperature during deposition changes the microstructure of the Be layer, leading to a significantly lower Ne retention. The empirical scaling equation was developed assuming that the Ne/Be ratio is proportional to the deposition rate of Be and the ion energy, obeying an Arrhenius-type temperature dependence. The resulting model accurately fits the experimental data obtained for the following conditions: substrate temperature 473 K ≤ T ≤ 573 K; deposition rate 0.04 nm/s ≤ r<sub>d</sub> ≤ 0.12 nm/s; ion energy 30 eV < <em>E<
{"title":"The study of neon retention and release behaviour from Be layers deposited under distinct temperature, pressure and ion energy conditions","authors":"P. Dinca , C. Staicu , B. Butoi , B.G. Solomonea , A. Anghel , O.G. Pompilian , G. Bulai , V. Tiron","doi":"10.1016/j.nme.2025.101877","DOIUrl":"10.1016/j.nme.2025.101877","url":null,"abstract":"<div><div>The purpose of this work is to carry out a systematic parametric study on neon (Ne) retention and release behaviour from Be/Ne layers co-deposited under ITER-relevant conditions, in terms of deposition temperature, gas pressure and ion energy. This study incorporates analyses of structural and morphological properties of Be/Ne coatings, along with Ne retention/release behaviour. A total of 27 batches of Be/Ne layers were deposited onto silicon (Si) and tungsten (W) substrates by adjusting to several selected values (<em>i</em>) the working gas pressure (1 Pa, 2 Pa, 3 Pa), (<em>ii</em>) the ion energy of the plasma species (30 eV, 100 eV, 200 eV) and (<em>iii</em>) the substrate temperature during deposition (340 K, 473 K, 573 K). The chosen deposition method was the novel Bipolar-High Power Impulse Magnetron Sputtering (BP-HiPIMS) technique due to its high ionization degree of the plasma species and its versatility in controlling the ion flux and energy during the deposition process. For a more comprehensive understanding, of how plasma properties, coatings’ structure, retention and release behaviour of Ne from the deposited samples correlate to each other, plasma-diagnosis was performed. The influence of gas pressure and pulsing configuration on ion energy and flux was investigated. The coating’s morphology and microstructure were analysed by Scanning Electron Microscopy (SEM), Atomic Force Microscopy (AFM) and X-ray diffraction (XRD). The surface’s morphology of the Be-Ne layers deposited at 340 K indicates smooth layers, with surface roughness values independent of the working gas pressure and ion energy. SEM images of the layers produced at temperatures higher than 340 K suggest the formation of blister-like structures on their surface. Unsurprisingly, the crystalline structure is strongly dependent on the substrate temperature. Metallic Be crystalline phase is observed for layers deposited at 473 K and 573 K, while, for the lowest substrate temperature (340 K), the structure of the Be-Ne layers was found to be amorphous, regardless of the working gas pressure and/or ion energy used during the deposition process. Ne/Be ratio in the deposited layers was evaluated through Thermal Desorption Spectroscopy (TDS) measurements. The results show an increase in the Ne content in layers with the increase of working gas pressure and ion energy. Although the Ne inventory is released above 1000 K, the increase of substrate temperature during deposition changes the microstructure of the Be layer, leading to a significantly lower Ne retention. The empirical scaling equation was developed assuming that the Ne/Be ratio is proportional to the deposition rate of Be and the ion energy, obeying an Arrhenius-type temperature dependence. The resulting model accurately fits the experimental data obtained for the following conditions: substrate temperature 473 K ≤ T ≤ 573 K; deposition rate 0.04 nm/s ≤ r<sub>d</sub> ≤ 0.12 nm/s; ion energy 30 eV < <em>E<","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101877"},"PeriodicalIF":2.3,"publicationDate":"2025-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173097","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}
The tritium inventory of the plasma-facing materials in a Large Helical Device (LHD) was analyzed using a thermal desorption system. In this study, we developed an induction heating system for the thermal desorption of tritium from carbon divertor tiles, which has several advantages, including a rapid temperature rise, internal heating via eddy currents, and heating with the original size and shape of the carbon tile. The apparatus was capable of heating a carbon divertor tile to a temperature greater than 1373 K over 20 min. Following a 12-h heating period at 1423 K, the tritium release rate from the carbon tile was 99.7 %. The chemical form of the tritium released from the divertor tiles was approximately 80 % molecular hydrogen, with the remaining tritium in the form of water vapor. The results of the tritium analysis of the divertor tiles suggest that the tritium distribution in the divertor region is not uniform, and that the flux varies from location to location in the LHD. To improve the accuracy of the tritium inventory analysis in an LHD vacuum vessel, it is essential to conduct a tritium inventory analysis of other plasma-facing materials.
{"title":"Determination of tritium inventory in carbon divertor tiles used in deuterium plasma experiment by induction heating method","authors":"Masahiro Tanaka , Hiromi Kato , Naoyuki Suzuki , Hiroki Chimura , Hiroaki Yonezu , Suguru Masuzaki","doi":"10.1016/j.nme.2025.101876","DOIUrl":"10.1016/j.nme.2025.101876","url":null,"abstract":"<div><div>The tritium inventory of the plasma-facing materials in a Large Helical Device (LHD) was analyzed using a thermal desorption system. In this study, we developed an induction heating system for the thermal desorption of tritium from carbon divertor tiles, which has several advantages, including a rapid temperature rise, internal heating via eddy currents, and heating with the original size and shape of the carbon tile. The apparatus was capable of heating a carbon divertor tile to a temperature greater than 1373 K over 20 min. Following a 12-h heating period at 1423 K, the tritium release rate from the carbon tile was 99.7 %. The chemical form of the tritium released from the divertor tiles was approximately 80 % molecular hydrogen, with the remaining tritium in the form of water vapor. The results of the tritium analysis of the divertor tiles suggest that the tritium distribution in the divertor region is not uniform, and that the flux varies from location to location in the LHD. To improve the accuracy of the tritium inventory analysis in an LHD vacuum vessel, it is essential to conduct a tritium inventory analysis of other plasma-facing materials.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101876"},"PeriodicalIF":2.3,"publicationDate":"2025-01-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173098","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 : 2025-01-17DOI: 10.1016/j.nme.2025.101866
X.Q. Xu , N.M. Li , M.L. Zhao , X. Liu , P.H. Diamond , B. Zhu , T.D. Rognlien , G.S. Xu
This study delves into the phenomena of fluctuation within the Scrape-Off Layer (SOL) of tokamak fusion reactors, with a specific focus on its impact on SOL width, particularly during grassy Edge Localized Modes (ELMs). Employing a comprehensive approach involving analysis and simulations, including BOUT++ and UEDGE simulations, we examine the role played by the fluctuation energy flux () and various control parameters in shaping the dynamics of fluctuation entrainment. Our findings shed light on the intricate process of transporting fluctuation energy from the pedestal to the SOL, a process influenced by critical factors such as fluctuation correlation length and SOL electric field shear. These insights yield valuable strategies for effectively managing SOL width and achieving plasma detachment, both of which are essential for optimizing fusion reactor performance and ensuring operational stability. Notably, the adoption of H-mode with small/grassy ELMs holds significant promise in addressing three pivotal challenges for future tokamak fusion reactors: minimizing the size of ELMs, expanding the SOL width, and aiding in the detachment of divertor plasma at lower plasma densities, all while avoiding back transitions to low confinement and disruptions.
{"title":"Fluctuation entrainment and SOL width broadening in small/grassy ELM regime","authors":"X.Q. Xu , N.M. Li , M.L. Zhao , X. Liu , P.H. Diamond , B. Zhu , T.D. Rognlien , G.S. Xu","doi":"10.1016/j.nme.2025.101866","DOIUrl":"10.1016/j.nme.2025.101866","url":null,"abstract":"<div><div>This study delves into the phenomena of fluctuation within the Scrape-Off Layer (SOL) of tokamak fusion reactors, with a specific focus on its impact on SOL width, particularly during grassy Edge Localized Modes (ELMs). Employing a comprehensive approach involving analysis and simulations, including BOUT++ and UEDGE simulations, we examine the role played by the fluctuation energy flux (<span><math><msub><mi>Γ</mi><mi>ε</mi></msub></math></span>) and various control parameters in shaping the dynamics of fluctuation entrainment. Our findings shed light on the intricate process of transporting fluctuation energy from the pedestal to the SOL, a process influenced by critical factors such as fluctuation correlation length and SOL electric field shear. These insights yield valuable strategies for effectively managing SOL width and achieving plasma detachment, both of which are essential for optimizing fusion reactor performance and ensuring operational stability. Notably, the adoption of H-mode with small/grassy ELMs holds significant promise in addressing three pivotal challenges for future tokamak fusion reactors: minimizing the size of ELMs, expanding the SOL width, and aiding in the detachment of divertor plasma at lower plasma densities, all while avoiding back transitions to low confinement and disruptions.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101866"},"PeriodicalIF":2.3,"publicationDate":"2025-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173566","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}
In this study, corrosion behavior of F82H reduced activation ferritic/martensitic steel (RAFM) by LTZO (Li2+xTiO3+y solid solution with 20 wt% Li2ZrO3) ceramic breeder pebbles was investigated at 773–998 K in an inert sweep gas condition (Ar + 0.1 % H2). Due to vapor gas release from the breeder pebbles and those penetrations, corrosion layer formed on the surface of the F82H steel. Glow discharge optical emission spectroscopy (GD-OES) and X-ray diffraction (XRD) identified the corrosion products as cubic, spinel, and rhombohedral Li–TM–O (TM: transition element in F82H such as Fe, Cr, and Mn). The growth of the corrosion layer thickness followed a parabolic curve at 833 K, yielding apparent diffusion coefficient of D = 6.95 × 10–13 cm2/s. Rapid growth was observed at 993 K after a parabolic growth which could be triggered by failure of the protective layer. A comparative analysis indicates a predominant effect of humidity and oxygen in the sweep gas on the growth rate, while the composition and shape of breeding materials have minor impacts.
{"title":"Corrosive behavior of structural F82H RAFM steel by LTZO ceramic breeder pebbles","authors":"Kosuke Kataoka , Keisuke Mukai , Juro Yagi , Motoki Nakajima , Jae-Hwan Kim , Takashi Nozawa","doi":"10.1016/j.nme.2025.101875","DOIUrl":"10.1016/j.nme.2025.101875","url":null,"abstract":"<div><div>In this study, corrosion behavior of F82H reduced activation ferritic/martensitic steel (RAFM) by LTZO (Li<sub>2+</sub><em><sub>x</sub></em>TiO<sub>3+</sub><em><sub>y</sub></em> solid solution with 20 wt% Li<sub>2</sub>ZrO<sub>3</sub>) ceramic breeder pebbles was investigated at 773–998 K in an inert sweep gas condition (Ar + 0.1 % H<sub>2</sub>). Due to vapor gas release from the breeder pebbles and those penetrations, corrosion layer formed on the surface of the F82H steel. Glow discharge optical emission spectroscopy (GD-OES) and X-ray diffraction (XRD) identified the corrosion products as cubic, spinel, and rhombohedral Li–TM–O (TM: transition element in F82H such as Fe, Cr, and Mn). The growth of the corrosion layer thickness followed a parabolic curve at 833 K, yielding apparent diffusion coefficient of <em>D</em> = 6.95 × 10<sup>–13</sup> cm<sup>2</sup>/s. Rapid growth was observed at 993 K after a parabolic growth which could be triggered by failure of the protective layer. A comparative analysis indicates a predominant effect of humidity and oxygen in the sweep gas on the growth rate, while the composition and shape of breeding materials have minor impacts.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101875"},"PeriodicalIF":2.3,"publicationDate":"2025-01-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173094","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 : 2025-01-14DOI: 10.1016/j.nme.2025.101867
D. Silvagni , O. Grover , A. Stagni , J.W. Hughes , M.A. Miller , B. Lomanowski , G. Ciraolo , M. Dunne , T. Eich , L. Frassinetti , C. Giroud , I. Jepu , A. Kallenbach , A. Kirjasuo , A. Kuang , T. Luda , C. Perez von Thun , T. Pütterich , H.J. Sun , H. Zohm
The separatrix electron density is an important parameter for core-edge scenario integration in tokamak devices, as it influences plasma confinement, divertor detachment and disruption avoidance. This quantity has been measured in H-mode discharges on JET, ASDEX Upgrade and Alcator C-Mod by applying the same fitting function to Thomson scattering measurements, and by employing the same analysis technique based on scrape-off layer power balance. To estimate the power crossing the separatrix, the inter-ELM time derivative of the plasma energy has been experimentally evaluated and found to be approximately a constant fraction of the absorbed heating power. Correlations between and engineering parameters have been investigated, revealing that scales with the divertor neutral pressure in a similar manner across all devices. Additionally, when is normalized to the obtained dependency, no clear correlation with the plasma current is found. These observations are in agreement with the 2-point model, which suggests that the upstream separatrix density is mainly set by the recycling at the divertor target.
{"title":"The separatrix electron density in JET, ASDEX upgrade and alcator C-Mod H-mode plasmas: A common evaluation procedure and correlation with engineering parameters","authors":"D. Silvagni , O. Grover , A. Stagni , J.W. Hughes , M.A. Miller , B. Lomanowski , G. Ciraolo , M. Dunne , T. Eich , L. Frassinetti , C. Giroud , I. Jepu , A. Kallenbach , A. Kirjasuo , A. Kuang , T. Luda , C. Perez von Thun , T. Pütterich , H.J. Sun , H. Zohm","doi":"10.1016/j.nme.2025.101867","DOIUrl":"10.1016/j.nme.2025.101867","url":null,"abstract":"<div><div>The separatrix electron density is an important parameter for core-edge scenario integration in tokamak devices, as it influences plasma confinement, divertor detachment and disruption avoidance. This quantity has been measured in H-mode discharges on JET, ASDEX Upgrade and Alcator C-Mod by applying the same fitting function to Thomson scattering measurements, and by employing the same analysis technique based on scrape-off layer power balance. To estimate the power crossing the separatrix, the inter-ELM time derivative of the plasma energy <span><math><mrow><mi>d</mi><mi>W</mi><mo>/</mo><mi>d</mi><mi>t</mi></mrow></math></span> has been experimentally evaluated and found to be approximately a constant fraction of the absorbed heating power. Correlations between <span><math><msub><mrow><mi>n</mi></mrow><mrow><mi>e</mi><mo>,</mo><mi>sep</mi></mrow></msub></math></span> and engineering parameters have been investigated, revealing that <span><math><msub><mrow><mi>n</mi></mrow><mrow><mi>e</mi><mo>,</mo><mi>sep</mi></mrow></msub></math></span> scales with the divertor neutral pressure <span><math><msub><mrow><mi>p</mi></mrow><mrow><mn>0</mn><mo>,</mo><mi>div</mi></mrow></msub></math></span> in a similar manner across all devices. Additionally, when <span><math><msub><mrow><mi>n</mi></mrow><mrow><mi>e</mi><mo>,</mo><mi>sep</mi></mrow></msub></math></span> is normalized to the obtained <span><math><msub><mrow><mi>p</mi></mrow><mrow><mn>0</mn><mo>,</mo><mi>div</mi></mrow></msub></math></span> dependency, no clear correlation with the plasma current is found. These observations are in agreement with the 2-point model, which suggests that the upstream separatrix density is mainly set by the recycling at the divertor target.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101867"},"PeriodicalIF":2.3,"publicationDate":"2025-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173572","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 : 2025-01-13DOI: 10.1016/j.nme.2025.101864
Pyry Virtanen , Henri Kumpulainen , Roni Mäenpää , Mathias Groth , Juri Romazanov , Sebastijan Brezinsek , JET contributors
Nickel transport in the Joint European Torus with the ITER-like wall (JET-ILW) is predicted using the 3D Monte-Carlo code ERO2.0, simulating the erosion and deposition of impurities in 3D geometry and utilizing hydrogenic background plasmas generated by the 2D edge fluid code EDGE2D-EIRENE. Charge exchange fluxes are obtained from the 3D neutral Monte-Carlo code EIRENE, which are modified to account for the shielding of the vacuum vessel wall by protruding plasma facing components, such as guard limiters. ERO2.0 is used to predict Ni erosion and deposition profiles for the first three JET-ILW campaigns weighted for the plasma operation time.
The primary location of nickel erosion on the Inconel vacuum vessel wall is predicted to be on the low-field side close to the midplane. The eroded nickel is predicted to be transported onto the entrance of the high-field side divertor, due to the scrape-off layer flows, where it is predicted to deposit and to form a layer on tile 1. The peak thickness of the predicted deposit layer is of the order 1-2/cm, a factor of six higher than measured in post-mortem tile analysis.
{"title":"ERO2.0 predictions of nickel migration in the JET ITER-Like Wall","authors":"Pyry Virtanen , Henri Kumpulainen , Roni Mäenpää , Mathias Groth , Juri Romazanov , Sebastijan Brezinsek , JET contributors","doi":"10.1016/j.nme.2025.101864","DOIUrl":"10.1016/j.nme.2025.101864","url":null,"abstract":"<div><div>Nickel transport in the Joint European Torus with the ITER-like wall (JET-ILW) is predicted using the 3D Monte-Carlo code ERO2.0, simulating the erosion and deposition of impurities in 3D geometry and utilizing hydrogenic background plasmas generated by the 2D edge fluid code EDGE2D-EIRENE. Charge exchange fluxes are obtained from the 3D neutral Monte-Carlo code EIRENE, which are modified to account for the shielding of the vacuum vessel wall by protruding plasma facing components, such as guard limiters. ERO2.0 is used to predict Ni erosion and deposition profiles for the first three JET-ILW campaigns weighted for the plasma operation time.</div><div>The primary location of nickel erosion on the Inconel vacuum vessel wall is predicted to be on the low-field side close to the midplane. The eroded nickel is predicted to be transported onto the entrance of the high-field side divertor, due to the scrape-off layer flows, where it is predicted to deposit and to form a layer on tile 1. The peak thickness of the predicted deposit layer is of the order 1-2<span><math><mrow><mi>⋅</mi><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>19</mn></mrow></msup></mrow></math></span>/cm<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span>, a factor of six higher than measured in post-mortem tile analysis.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101864"},"PeriodicalIF":2.3,"publicationDate":"2025-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173088","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 : 2025-01-13DOI: 10.1016/j.nme.2025.101869
W. Xu , Z. Wang , Z. Sun , R. Maingi , Z.T. Zhou , Y.H. Guan , Y. Zhu , X.C. Meng , M. Huang , Y.W. Yu , G.Z. Zuo , J.S. Hu
The limit of boron flow rates for real-time conditioning of the first walls has been systematically investigated in the Experimental Advanced Superconducting Tokamak (EAST) with a full metal wall. Initially, solid boron injection demonstrated effective control over carbon impurities and deuterium recycling on the basis of pre-discharge boronization. A minimum flow rate, identified between 1.0 mg/s and 2.0 mg/s, was necessary for actively improving wall conditions under specific plasma operating scenarios, with this effect progressively enhancing as boron flow rates increased. Additionally, a maximum flow rate, estimated between 3.5 mg/s and 8.0 mg/s, was identified for these plasma conditions. When boron flow rates exceeded this maximum, boron-induced fueling effects influenced the plasma line-averaged density, and at excessively high flow rates, plasma disruption was observed.
{"title":"Investigation of boron powder flow rates on real-time wall","authors":"W. Xu , Z. Wang , Z. Sun , R. Maingi , Z.T. Zhou , Y.H. Guan , Y. Zhu , X.C. Meng , M. Huang , Y.W. Yu , G.Z. Zuo , J.S. Hu","doi":"10.1016/j.nme.2025.101869","DOIUrl":"10.1016/j.nme.2025.101869","url":null,"abstract":"<div><div>The limit of boron flow rates for real-time conditioning of the first walls has been systematically investigated in the Experimental Advanced Superconducting Tokamak (EAST) with a full metal wall. Initially, solid boron injection demonstrated effective control over carbon impurities and deuterium recycling on the basis of pre-discharge boronization. A minimum flow rate, identified between 1.0 mg/s and 2.0 mg/s, was necessary for actively improving wall conditions under specific plasma operating scenarios, with this effect progressively enhancing as boron flow rates increased. Additionally, a maximum flow rate, estimated between 3.5 mg/s and 8.0 mg/s, was identified for these plasma conditions. When boron flow rates exceeded this maximum, boron-induced fueling effects influenced the plasma line-averaged density, and at excessively high flow rates, plasma disruption was observed.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101869"},"PeriodicalIF":2.3,"publicationDate":"2025-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173091","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 : 2025-01-13DOI: 10.1016/j.nme.2025.101868
B.J. Peterson , G. Partesotti , F. Reimold , G.A. Wurden , Y. Gao , D. Zhang , V. Winters , M. Kobayashi , Y. Feng , K. Mukai , J. von Miller , the W7-X Team
Mitigation of heat on the first wall through divertor operation is a key to a successful future fusion reactor. W7-X employs an island divertor to control the exhaust and heat load on the plasma impacting divertor plates. Increased radiation in the divertor reduces the heat load at the plasma contact point during detachment. In this paper we investigate the distribution of the radiation using an InfraRed imaging Video Bolometer (IRVB) that views the divertor region in two dimensions giving information on both the poloidal and toroidal variation of the radiation in comparison to conventional resistive bolometer arrays that typically only give poloidal variation information. Experiments were carried out using a standard magnetic configuration modified by changing control and planar coil currents to achieve three different island sizes without changing the strike line location. For each island size low and high density (ne = ∼4 and ∼ 7 x 1019/m3, respectively) plasmas were created with ∼ 2 MW of ECH input power, which correspond to attached and detached plasmas with radiated power fractions (frad) of ∼ 20–25 % and ∼ 90 %, respectively.
Results indicate an increase in density led to an increase in the IRVB radiation signals as seen in the total radiated power (and frad) and a slight broadening in the signals indicating less radiation from the target locations, especially the lower right location in the IRVB field of view when compared with the corresponding thermography images. However, no noticeable difference in the IRVB radiation pattern or intensity is seen with the change of the island size.
{"title":"Investigation of island size effect on radiation distribution during attached and detached plasmas in the island divertor of W7-X","authors":"B.J. Peterson , G. Partesotti , F. Reimold , G.A. Wurden , Y. Gao , D. Zhang , V. Winters , M. Kobayashi , Y. Feng , K. Mukai , J. von Miller , the W7-X Team","doi":"10.1016/j.nme.2025.101868","DOIUrl":"10.1016/j.nme.2025.101868","url":null,"abstract":"<div><div>Mitigation of heat on the first wall through divertor operation is a key to a successful future fusion reactor. W7-X employs an island divertor to control the exhaust and heat load on the plasma impacting divertor plates. Increased radiation in the divertor reduces the heat load at the plasma contact point during detachment. In this paper we investigate the distribution of the radiation using an InfraRed imaging Video Bolometer (IRVB) that views the divertor region in two dimensions giving information on both the poloidal and toroidal variation of the radiation in comparison to conventional resistive bolometer arrays that typically only give poloidal variation information. Experiments were carried out using a standard magnetic configuration modified by changing control and planar coil currents to achieve three different island sizes without changing the strike line location. For each island size low and high density (n<sub>e</sub> = ∼4 and ∼ 7 x 10<sup>19</sup>/m<sup>3</sup>, respectively) plasmas were created with ∼ 2 MW of ECH input power, which correspond to attached and detached plasmas with radiated power fractions (f<sub>rad</sub>) of ∼ 20–25 % and ∼ 90 %, respectively.</div><div>Results indicate an increase in density led to an increase in the IRVB radiation signals as seen in the total radiated power (and f<sub>rad</sub>) and a slight broadening in the signals indicating less radiation from the target locations, especially the lower right location in the IRVB field of view when compared with the corresponding thermography images. However, no noticeable difference in the IRVB radiation pattern or intensity is seen with the change of the island size.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101868"},"PeriodicalIF":2.3,"publicationDate":"2025-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173092","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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