Pub Date : 2024-12-01DOI: 10.1016/j.nme.2024.101819
Thomas Body , Thomas Eich , Adam Kuang , Tom Looby , Mike Kryjak , Ben Dudson , Matthew Reinke
Fusion power plants will require detachment to mitigate sputtering and keep divertor heat fluxes at tolerable levels. Controlling detachment on these devices may require the use of real-time scrape-off-layer modeling to complement the limited set of available diagnostics. In this work, we use the configurable Hermes-3 edge modeling framework to perform time-dependent, fixed-fraction-impurity 1D detachment simulations. Although currently far from real-time, these simulations are used to investigate time-dependent effects and the minimum physics set required for control-relevant modeling. We show that these simulations reproduce the expected rollover of the target ion flux — a typical characteristic of detachment onset. We also perform scans of the input heat flux and impurity concentration and show that the steady-state results closely match the scalings predicted by the 0D time-independent Lengyel–Goedheer model. This allows us to indirectly compare to SOLPS simulations, which find a similar scaling but a lower value for the impurity concentration required for detachment for given upstream conditions. We use this result to suggest a series of improvements for the Hermes simulations, and finally show simulations demonstrating the impact of time-dependence.
{"title":"Detachment scalings derived from 1D scrape-off-layer simulations","authors":"Thomas Body , Thomas Eich , Adam Kuang , Tom Looby , Mike Kryjak , Ben Dudson , Matthew Reinke","doi":"10.1016/j.nme.2024.101819","DOIUrl":"10.1016/j.nme.2024.101819","url":null,"abstract":"<div><div>Fusion power plants will require detachment to mitigate sputtering and keep divertor heat fluxes at tolerable levels. Controlling detachment on these devices may require the use of real-time scrape-off-layer modeling to complement the limited set of available diagnostics. In this work, we use the configurable Hermes-3 edge modeling framework to perform time-dependent, fixed-fraction-impurity 1D detachment simulations. Although currently far from real-time, these simulations are used to investigate time-dependent effects and the minimum physics set required for control-relevant modeling. We show that these simulations reproduce the expected rollover of the target ion flux — a typical characteristic of detachment onset. We also perform scans of the input heat flux and impurity concentration and show that the steady-state results closely match the scalings predicted by the 0D time-independent Lengyel–Goedheer model. This allows us to indirectly compare to SOLPS simulations, which find a similar scaling but a lower value for the impurity concentration required for detachment for given upstream conditions. We use this result to suggest a series of improvements for the Hermes simulations, and finally show simulations demonstrating the impact of time-dependence.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101819"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142743334","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 : 2024-12-01DOI: 10.1016/j.nme.2024.101811
M. Zhao, F. Scotti, T.D. Rognlien, A.G. McLean, G. Burke, A. Holm
UEDGE simulations with density scans for various input power, transport coefficients and outer poloidal leg length are performed to study the conditions for the existence of a bifurcation-like drop of at the outer strike point, commonly referred to as a detachment cliff, when transitioning to a detached plasma from an attached plasma in the outer divertor as the upstream density increases (McLean et al., 2015). The simulation results show that a detachment cliff tends to occur with a higher power input regardless of diffusivities and leg length. Further analysis of change of plasma profiles at a cliff indicate that, in addition to the sharp reduction of the drift fluxes in the outer divertor studied in Jaervinen et al., (2018), the substantial change of the Mach number in the outer divertor and the decrease of the outer mid-plane due to the radiation front moving across the separatrix into the confinement region above the X-point consistently occur for all UEDGE density scans that have a detachment cliff. UEDGE time-dependent simulation of the evolution of a detachment cliff shows that the rapid increase of radiation above the X-point occurs in a time scale of , which could possibly be the trigger for the formation of a detachment cliff, quicker than the Mach number change in a time scale of and the drop of in a time scale of in the outer divertor.
UEDGE模拟对各种输入功率、输运系数和外部极向支腿长度进行了密度扫描，以研究随着上游密度的增加，当外部分流器中的附着等离子体过渡到分离等离子体时，在外部击点（通常称为分离悬崖）存在分叉状Te滴的条件（McLean et al., 2015）。仿真结果表明，无论扩散系数和支腿长度如何，在较高的功率输入下，往往会出现分离悬崖。对悬崖处等离子体剖面变化的进一步分析表明，除了Jaervinen等人（2018）所研究的外导流器E×B漂移通量急剧减少之外，由于辐射锋穿过分离矩阵进入x点以上的约束区域，所有具有分离悬崖的UEDGE密度扫描都一致发生了外导流器马赫数的实质性变化和外中平面Te的减少。UEDGE对分离悬崖演化的时间依赖模拟表明，x点以上辐射的快速增加发生在~ 0.3-0.5ms的时间尺度上，这可能是分离悬崖形成的触发因素，比外导流器马赫数的变化在~ 1ms的时间尺度上和Te的下降在~ 2-3ms的时间尺度上要快。
{"title":"2D analysis of tokamak divertor-plasma detachment-bifurcation with operational parameters and geometries","authors":"M. Zhao, F. Scotti, T.D. Rognlien, A.G. McLean, G. Burke, A. Holm","doi":"10.1016/j.nme.2024.101811","DOIUrl":"10.1016/j.nme.2024.101811","url":null,"abstract":"<div><div>UEDGE simulations with density scans for various input power, transport coefficients and outer poloidal leg length are performed to study the conditions for the existence of a bifurcation-like drop of <span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span> at the outer strike point, commonly referred to as a detachment cliff, when transitioning to a detached plasma from an attached plasma in the outer divertor as the upstream density increases (McLean et al., 2015). The simulation results show that a detachment cliff tends to occur with a higher power input regardless of diffusivities and leg length. Further analysis of change of plasma profiles at a cliff indicate that, in addition to the sharp reduction of the <span><math><mrow><mi>E</mi><mo>×</mo><mi>B</mi></mrow></math></span> drift fluxes in the outer divertor studied in Jaervinen et al., (2018), the substantial change of the Mach number in the outer divertor and the decrease of the outer mid-plane <span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span> due to the radiation front moving across the separatrix into the confinement region above the X-point consistently occur for all UEDGE density scans that have a detachment cliff. UEDGE time-dependent simulation of the evolution of a detachment cliff shows that the rapid increase of radiation above the X-point occurs in a time scale of <span><math><mrow><mo>∼</mo><mn>0</mn><mo>.</mo><mn>3</mn><mtext>–</mtext><mn>0</mn><mo>.</mo><mn>5</mn><mspace></mspace><mi>ms</mi></mrow></math></span>, which could possibly be the trigger for the formation of a detachment cliff, quicker than the Mach number change in a time scale of <span><math><mrow><mo>∼</mo><mn>1</mn><mspace></mspace><mi>ms</mi></mrow></math></span> and the drop of <span><math><msub><mrow><mi>T</mi></mrow><mrow><mi>e</mi></mrow></msub></math></span> in a time scale of <span><math><mrow><mo>∼</mo><mn>2</mn><mtext>–</mtext><mn>3</mn><mspace></mspace><mi>ms</mi></mrow></math></span> in the outer divertor.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101811"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142743335","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 : 2024-12-01DOI: 10.1016/j.nme.2024.101840
Qian Long , Guoliang Xu , Rui Ding , Hui Wang , Yunjia Zhang , Ran Hai , Xue Bai , Bingfu Gao , Junling Chen
Lithium (Li) coating of the first wall has been used as a routine way of wall conditioning on EAST and proved to have a significant influence on particle recycling. Whereas the deuterium (D) reflection process on tungsten (W) with Li coating is not well understood. The SDTrim.SP code is employed to investigate the D reflection process on the Li-W mixed material wall surface for EAST tokamak. The study reveals that the existence of Li coating can reduce the deuterium reflection rate significantly. This reduction in reflection is influenced by both the D impact energy and the thickness of the Li overlayer. Unlike the D reflection on pure metal materials, the D reflection rate shows a non-monotonic relationship with the D impact energy. Detail analysis shows that with the increase of D impact energy, the fraction of D which can penetrate through the Li overlayer and interact with the W substrate increases, leading to a significant increase of D reflection rate. For normal EAST divertor operation conditions, a Li overlayer thickness of 30 nm is required to keep the D reflection rate at a low level. The D reflection on W with a Li-W mixed overlayer has also been investigated and discussed in this work.
{"title":"Study of deuterium reflection on tungsten material with lithium coating in EAST tokamak","authors":"Qian Long , Guoliang Xu , Rui Ding , Hui Wang , Yunjia Zhang , Ran Hai , Xue Bai , Bingfu Gao , Junling Chen","doi":"10.1016/j.nme.2024.101840","DOIUrl":"10.1016/j.nme.2024.101840","url":null,"abstract":"<div><div>Lithium (Li) coating of the first wall has been used as a routine way of wall conditioning on EAST and proved to have a significant influence on particle recycling. Whereas the deuterium (D) reflection process on tungsten (W) with Li coating is not well understood. The SDTrim.SP code is employed to investigate the D reflection process on the Li-W mixed material wall surface for EAST tokamak. The study reveals that the existence of Li coating can reduce the deuterium reflection rate significantly. This reduction in reflection is influenced by both the D impact energy and the thickness of the Li overlayer. Unlike the D reflection on pure metal materials, the D reflection rate shows a non-monotonic relationship with the D impact energy. Detail analysis shows that with the increase of D impact energy, the fraction of D which can penetrate through the Li overlayer and interact with the W substrate increases, leading to a significant increase of D reflection rate. For normal EAST divertor operation conditions, a Li overlayer thickness of 30 nm is required to keep the D reflection rate at a low level. The D reflection on W with a Li-W mixed overlayer has also been investigated and discussed in this work.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101840"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143156749","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 : 2024-12-01DOI: 10.1016/j.nme.2024.101834
Yaowei Yu, Hao Sun, Chao Wang, Bin Cao, Guizhong Zuo, Jiansheng Hu
Fuel (deuterium) removal by various helium (He) discharge cleanings under strong magnetic field are studied in EAST superconducting tokamak, including deuterium-to-helium changeover by main plasma operation, glow discharge cleaning (GDC), and ion cyclotron wall conditioning (ICWC). The study demonstrates that He-GDC works well in strong magnetic field, but the cleaning is poloidally located near the GDC anodes. Both He-GDC and He-ICWC are effective in removing deuterium under strong magnetic field. Moreover, deuterium-to-helium changeover by main plasma operation is highly effective in rapidly reducing the deuterium from the first wall surface. A comprehensive comparison of the three techniques shows that deuterium-to-helium changeover and He-GDC exhibit higher deuterium removal rate, and the pulse duration and duty cycle of He-ICWC should be optimized to improve the deuterium removal. These studies present a comprehensive investigation on fuel removal techniques in EAST tokamak, providing valuable insights into the optimization of tritium removal strategies for future fusion reactors.
{"title":"Removal of deuterium retention by various helium discharge cleanings under strong magnetic field in EAST superconducting tokamak","authors":"Yaowei Yu, Hao Sun, Chao Wang, Bin Cao, Guizhong Zuo, Jiansheng Hu","doi":"10.1016/j.nme.2024.101834","DOIUrl":"10.1016/j.nme.2024.101834","url":null,"abstract":"<div><div>Fuel (deuterium) removal by various helium (He) discharge cleanings under strong magnetic field are studied in EAST superconducting tokamak, including deuterium-to-helium changeover by main plasma operation, glow discharge cleaning (GDC), and ion cyclotron wall conditioning (ICWC). The study demonstrates that He-GDC works well in strong magnetic field, but the cleaning is poloidally located near the GDC anodes. Both He-GDC and He-ICWC are effective in removing deuterium under strong magnetic field. Moreover, deuterium-to-helium changeover by main plasma operation is highly effective in rapidly reducing the deuterium from the first wall surface. A comprehensive comparison of the three techniques shows that deuterium-to-helium changeover and He-GDC exhibit higher deuterium removal rate, and the pulse duration and duty cycle of He-ICWC should be optimized to improve the deuterium removal. These studies present a comprehensive investigation on fuel removal techniques in EAST tokamak, providing valuable insights into the optimization of tritium removal strategies for future fusion reactors.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101834"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143156872","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 : 2024-12-01DOI: 10.1016/j.nme.2024.101825
T. Kulsartov , I. Kenzhina , Yu Ponkratov , Yu Gordienko , Zh Zaurbekova , K. Samarkhanov , S. Askerbekov , Ye A. Kenzhin , A.B. Yelishenkov
The SnLi alloy has good prospects to be used as a material of intra-chamber elements of fusion facilities, as it has a number of advantages over pure lithium. The results of a series of experiments on deuterium adsorption/desorption by a Sn73Li27 alloy sample are presented. To prepare an alloy sample with the required tin and lithium content, a technique was developed and a special experimental device was constructed. An ampoule device was manufactured to conduct a series of experiments to study sample saturation with deuterium and desorption of deuterium and deuterium-containing molecules from it. Saturation was carried out at alloy temperatures of 650, 600, 550, 500 and 450 °C. TDS experiments were carried out at 20 °C/min. The possible mechanism of deuterium dissolution and release from the tin-lithium alloy was considered and temperature dependences of the effective deuterium solubility constant KS in the tin-lithium alloy were calculated within the framework of the proposed mechanism. The temperature dependence of the Sieverts’ constant for the test sample in the temperature range of 500–650 °C was determined as
{"title":"Investigation of the interaction of deuterium with Sn73Li27 tin-lithium alloy","authors":"T. Kulsartov , I. Kenzhina , Yu Ponkratov , Yu Gordienko , Zh Zaurbekova , K. Samarkhanov , S. Askerbekov , Ye A. Kenzhin , A.B. Yelishenkov","doi":"10.1016/j.nme.2024.101825","DOIUrl":"10.1016/j.nme.2024.101825","url":null,"abstract":"<div><div>The SnLi alloy has good prospects to be used as a material of intra-chamber elements of fusion facilities, as it has a number of advantages over pure lithium. The results of a series of experiments on deuterium adsorption/desorption by a Sn<sub>73</sub>Li<sub>27</sub> alloy sample are presented. To prepare an alloy sample with the required tin and lithium content, a technique was developed and a special experimental device was constructed. An ampoule device was manufactured to conduct a series of experiments to study sample saturation with deuterium and desorption of deuterium and deuterium-containing molecules from it. Saturation was carried out at alloy temperatures of 650, 600, 550, 500 and 450 °C. TDS experiments were carried out at 20 °C/min. The possible mechanism of deuterium dissolution and release from the tin-lithium alloy was considered and temperature dependences of the effective deuterium solubility constant <em>K<sub>S</sub></em> in the tin-lithium alloy were calculated within the framework of the proposed mechanism. The temperature dependence of the Sieverts’ constant for the test sample in the temperature range of 500–650 °C was determined as <span><math><mrow><msub><mi>K</mi><mi>S</mi></msub><mfenced><mrow><mi>T</mi></mrow></mfenced><mo>=</mo><mn>3.6</mn><mo>·</mo><msup><mrow><mn>10</mn></mrow><mn>6</mn></msup><mo>∙</mo><mi>exp</mi><mfenced><mrow><mo>-</mo><mfrac><mrow><mn>32400</mn><mo>(</mo><mi>J</mi><mo>)</mo></mrow><mrow><mi>RT</mi></mrow></mfrac></mrow></mfenced><mo>∙</mo><msup><mrow><mi>Pa</mi></mrow><mfrac><mn>1</mn><mn>2</mn></mfrac></msup><mo>.</mo></mrow></math></span></div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101825"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142743339","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 : 2024-12-01DOI: 10.1016/j.nme.2024.101823
G. Partesotti , F. Reimold , G.A. Wurden , B.J. Peterson , D. Zhang , K. Mukai , the W7-X team
Radiative cooling is one of the main heat dissipation channels for magnetically confined fusion plasmas. In stellarators power exhaust control is complicated by the three-dimensional radiation distribution. Using a Gaussian Process Tomography routine, the two-dimensional radiation distribution in Wendelstein 7-X is reconstructed from the line-integrated bolometer data of one poloidal cross-section. These experimental tomograms are mapped to the field of view of the newly installed divertor bolometer cameras and compared to the toroidally separated local measurements to assess the toroidal radiation distribution. We observe peaking of the plasma radiated power density in the divertor region and steep toroidal gradients, especially at low radiated power fraction (up to increase in the emissivity along a m long flux tube). These results suggest the presence of significant toroidal radiation asymmetries within a stellarator half-module.
{"title":"Assessing the toroidal radiation distribution at Wendelstein 7-X by combining Gaussian Process Tomography and field line mapping","authors":"G. Partesotti , F. Reimold , G.A. Wurden , B.J. Peterson , D. Zhang , K. Mukai , the W7-X team","doi":"10.1016/j.nme.2024.101823","DOIUrl":"10.1016/j.nme.2024.101823","url":null,"abstract":"<div><div>Radiative cooling is one of the main heat dissipation channels for magnetically confined fusion plasmas. In stellarators power exhaust control is complicated by the three-dimensional radiation distribution. Using a Gaussian Process Tomography routine, the two-dimensional radiation distribution in Wendelstein 7-X is reconstructed from the line-integrated bolometer data of one poloidal cross-section. These experimental tomograms are mapped to the field of view of the newly installed divertor bolometer cameras and compared to the toroidally separated local measurements to assess the toroidal radiation distribution. We observe peaking of the plasma radiated power density in the divertor region and steep toroidal gradients, especially at low radiated power fraction (up to <span><math><mrow><mo>×</mo><mn>8</mn></mrow></math></span> increase in the emissivity along a <span><math><mrow><mo>∼</mo><mn>2</mn><mspace></mspace></mrow></math></span>m long flux tube). These results suggest the presence of significant toroidal radiation asymmetries within a stellarator half-module.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101823"},"PeriodicalIF":2.3,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143156871","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 : 2024-11-30DOI: 10.1016/j.nme.2024.101831
L. Colas , W. Helou , G. Urbanczyk , V. Bobkov , F. Calarco , N. Fedorczak , D. Milanesio , J. Hillairet
In 2023, switching the material on the first wall of ITER to tungsten (W) was recommended. In magnetic Fusion devices, waves in the Ion Cyclotron Range of Frequencies (ICRF) interact with the Scrape-Off Layer (SOL) via RF-sheath rectification. This contribution re-assesses this phenomenon close to the ITER ICRF antenna, focusing on the ICRF-specific gross erosion of W from the antenna port sides. Our quantitative estimates rely on predictive multi-2D numerical simulations of the ICRF antenna environment using the SSWICH-SW code. They combine Slow Wave propagation from the antenna mouth to the SOL, the excitation of RF oscillations in the sheath voltages at the antenna port sides and a subsequent DC biasing of the SOL. Maps of the parallel RF electric field at the antenna mouth, from the antenna code TOPICA, excite the system. Our simulations cover more than four decades in the local densities near the antenna. Since both the sputtering and the local heat loads are proportional to the local particle fluxes, the most intense Plasma-Wall Interaction is found for high local density, with or without ICRF waves. In these conditions, larger margins also exist for coupling the ICRF power. We tested several operational trade-offs between these two constraints. The simulated target plasma contains 2% of neon ions. These are efficient at sputtering W, already at low accelerating voltages. Consequently, although the RF-sheath rectification sufficiently amplifies the local sputtering at the antenna port for a detection using visible spectroscopy, the ICRF-induced increment of the gross W production represents at worse 22% of the W source expected from thermal sheaths over the eighteen out-board mid-plane ports. An upper bound, independent of our main assumptions, is proposed for this enhancement factor. This moderate expected global increase questions the ability to detect ICRF-specific W contamination of the plasma core, even at the planned maximal ICRF power.
{"title":"Numerical assessment of ICRF-specific plasma-wall interaction in the new ITER baseline using the SSWICH-SW code","authors":"L. Colas , W. Helou , G. Urbanczyk , V. Bobkov , F. Calarco , N. Fedorczak , D. Milanesio , J. Hillairet","doi":"10.1016/j.nme.2024.101831","DOIUrl":"10.1016/j.nme.2024.101831","url":null,"abstract":"<div><div>In 2023, switching the material on the first wall of ITER to tungsten (W) was recommended. In magnetic Fusion devices, waves in the Ion Cyclotron Range of Frequencies (ICRF) interact with the Scrape-Off Layer (SOL) <em>via</em> RF-sheath rectification. This contribution re-assesses this phenomenon close to the ITER ICRF antenna, focusing on the ICRF-specific gross erosion of W from the antenna port sides. Our quantitative estimates rely on predictive multi-2D numerical simulations of the ICRF antenna environment using the SSWICH-SW code. They combine Slow Wave propagation from the antenna mouth to the SOL, the excitation of RF oscillations in the sheath voltages at the antenna port sides and a subsequent DC biasing of the SOL. Maps of the parallel RF electric field at the antenna mouth, from the antenna code TOPICA, excite the system. Our simulations cover more than four decades in the local densities near the antenna. Since both the sputtering and the local heat loads are proportional to the local particle fluxes, the most intense Plasma-Wall Interaction is found for high local density, with or without ICRF waves. In these conditions, larger margins also exist for coupling the ICRF power. We tested several operational trade-offs between these two constraints. The simulated target plasma contains 2% of neon ions. These are efficient at sputtering W, already at low accelerating voltages. Consequently, although the RF-sheath rectification sufficiently amplifies the local sputtering at the antenna port for a detection using visible spectroscopy, the ICRF-induced increment of the gross W production represents at worse 22% of the W source expected from thermal sheaths over the eighteen out-board mid-plane ports. An upper bound, independent of our main assumptions, is proposed for this enhancement factor. This moderate expected global increase questions the ability to detect ICRF-specific W contamination of the plasma core, even at the planned maximal ICRF power.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"42 ","pages":"Article 101831"},"PeriodicalIF":2.3,"publicationDate":"2024-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143173559","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 : 2024-11-27DOI: 10.1016/j.nme.2024.101828
Wei Zheng , Rong Yan , Rui Ding , Guoliang Xu , Lei Mu , Yefan Zhu , Yuming Liu , Junlin Wan , Junling Chen
Plasma-wall interaction (PWI) is a critical concern in tokamaks because of its significant impact on the lifetimes of plasma-facing materials (PFMs), fuel retention, and plasma performance. In 2020, the Experimental Advanced Superconducting Tokamak (EAST) PFMs were upgraded to predominantly metallic walls. Following the 2021 experimental campaign, the deposition distribution and fuel retention on the surfaces of the PFMs along both the poloidal and toroidal directions were analyzed. The poloidal tests commenced at the high-field side (HFS), proceeded to the lower divertor, then to the low-field side (LFS), and finally to the upper divertor. The toroidal tests were performed at the midplane of the HFS, beginning from port A and ending at port P. The distributions of the deposits in the poloidal and toroidal directions were clearly asymmetrical. Mo and Fe particles sputtered from the first wall and inner stainless steel (SS) components were prone to deposition in the lower divertor region, as evidenced by the fact that the elemental content in the far-SOL region of the inner divertor exhibited Mo and Fe peaks, as did both the near- and far-SOL regions of the outer divertor. In addition, quick re-deposition of W and Fe was observed, as demonstrated by the fact that their contents near the erosion sources were higher than those farther away along the toroidal first wall on the HFS. This tendency was stronger for W than for Fe. Further analysis indicated that the deposits consisted of Li, C, O, W, Mo, Fe, Cu, and Ni. The Li and Fe contents were much higher than those of other metal impurities, with peak values of 8.17 μg/mm2 and 7.78 μg/mm2, respectively. The Li content decreased along the HFS first wall from 8.17 μg/mm2 at P-2 to 1.82 μg/mm2 at P-22, and then increased to 2.72 μg/mm2 at P-32 in the divertor, while the Fe content was higher around the top side and the midplane of the HFS. The deposited Li originated from routine wall conditioning and existed primarily in the form of Li2CO3. Additionally, other metal impurities deposited on the Li2CO3 surfaces exhibited various irregular shapes, often appearing as aggregated and recrystallized small particles. Furthermore, significant deuterium (D) retention on the order of 1021 atoms/m2 was measured on all the analyzed SiC-coated graphite tiles on the HFS. The D content at location P-10 at the midplane was the highest along the poloidal direction of the HFS because it was directly facing the NBI beam.
等离子体壁相互作用（PWI）对等离子体面材料（PFMs）的寿命、燃料保持和等离子体性能有重大影响，因此是托卡马克的一个关键问题。2020 年，先进超导托卡马克实验装置（EAST）的 PFMs 升级为主要为金属壁。在 2021 年的实验活动之后，对 PFM 表面沿极坐标和环坐标方向的沉积分布和燃料保留情况进行了分析。极环形试验从高场侧（HFS）开始，到下分流器，再到低场侧（LFS），最后到上分流器。环形试验在 HFS 的中平面进行，从端口 A 开始，到端口 P 结束。从第一壁和内部不锈钢（SS）部件溅射出的钼和铁颗粒容易沉积在分流器下部区域，这一点可以从内部分流器远太阳区域的元素含量显示钼和铁峰值，以及外部分流器近太阳和远太阳区域的元素含量显示钼和铁峰值得到证明。此外，还观察到 W 和 Fe 的快速再沉积，这表现在它们在侵蚀源附近的含量高于沿 HFS 环形第一壁较远处的含量。这种趋势在 W 方面比在 Fe 方面更为明显。进一步分析表明，沉积物由 Li、C、O、W、Mo、Fe、Cu 和 Ni 组成。锂和铁的含量远高于其他金属杂质，峰值分别为 8.17 μg/mm2 和 7.78 μg/mm2。沿 HFS 第一壁的锂含量从 P-2 时的 8.17 μg/mm2 降至 P-22 时的 1.82 μg/mm2，然后在分流器内 P-32 时增至 2.72 μg/mm2，而在 HFS 顶面和中面周围的铁含量较高。沉积的锂来自常规壁面调节，主要以 Li2CO3 的形式存在。此外，沉积在 Li2CO3 表面的其他金属杂质呈现出各种不规则形状，通常表现为聚集和再结晶的小颗粒。此外，在 HFS 上所有分析过的碳化硅涂层石墨瓦片上都测量到了大量的氘 (D) 保留，数量级为 1021 个原子/平方米。位于中平面 P-10 位置的 D 含量沿 HFS 的极坐标方向最高，因为该位置直接面对 NBI 光束。
{"title":"Post-mortem analysis of material deposition and fuel retention on the plasma-facing materials after the 2021 campaign in EAST","authors":"Wei Zheng , Rong Yan , Rui Ding , Guoliang Xu , Lei Mu , Yefan Zhu , Yuming Liu , Junlin Wan , Junling Chen","doi":"10.1016/j.nme.2024.101828","DOIUrl":"10.1016/j.nme.2024.101828","url":null,"abstract":"<div><div>Plasma-wall interaction (PWI) is a critical concern in tokamaks because of its significant impact on the lifetimes of plasma-facing materials (PFMs), fuel retention, and plasma performance. In 2020, the Experimental Advanced Superconducting Tokamak (EAST) PFMs were upgraded to predominantly metallic walls. Following the 2021 experimental campaign, the deposition distribution and fuel retention on the surfaces of the PFMs along both the poloidal and toroidal directions were analyzed. The poloidal tests commenced at the high-field side (HFS), proceeded to the lower divertor, then to the low-field side (LFS), and finally to the upper divertor. The toroidal tests were performed at the midplane of the HFS, beginning from port A and ending at port P. The distributions of the deposits in the poloidal and toroidal directions were clearly asymmetrical. Mo and Fe particles sputtered from the first wall and inner stainless steel (SS) components were prone to deposition in the lower divertor region, as evidenced by the fact that the elemental content in the far-SOL region of the inner divertor exhibited Mo and Fe peaks, as did both the near- and far-SOL regions of the outer divertor. In addition, quick re-deposition of W and Fe was observed, as demonstrated by the fact that their contents near the erosion sources were higher than those farther away along the toroidal first wall on the HFS. This tendency was stronger for W than for Fe. Further analysis indicated that the deposits consisted of Li, C, O, W, Mo, Fe, Cu, and Ni. The Li and Fe contents were much higher than those of other metal impurities, with peak values of 8.17 μg/mm<sup>2</sup> and 7.78 μg/mm<sup>2</sup>, respectively. The Li content decreased along the HFS first wall from 8.17 μg/mm<sup>2</sup> at P-2 to 1.82 μg/mm<sup>2</sup> at P-22, and then increased to 2.72 μg/mm<sup>2</sup> at P-32 in the divertor, while the Fe content was higher around the top side and the midplane of the HFS. The deposited Li originated from routine wall conditioning and existed primarily in the form of Li<sub>2</sub>CO<sub>3</sub>. Additionally, other metal impurities deposited on the Li<sub>2</sub>CO<sub>3</sub> surfaces exhibited various irregular shapes, often appearing as aggregated and recrystallized small particles. Furthermore, significant deuterium (D) retention on the order of 10<sup>21</sup> atoms/m<sup>2</sup> was measured on all the analyzed SiC-coated graphite tiles on the HFS. The D content at location P-10 at the midplane was the highest along the poloidal direction of the HFS because it was directly facing the NBI beam.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101828"},"PeriodicalIF":2.3,"publicationDate":"2024-11-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142723785","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 : 2024-11-23DOI: 10.1016/j.nme.2024.101822
Wenmin Zhang , Ling Zhang , Shigeru Morita , Yunxin Cheng , Hui Sheng , Chengxi Zhou , Huihui Wang , Youwen Sun , Yuqi Chu , Ning Sun , Ailan Hu , Darío Mitnik , Yinxian Jie , Haiqing Liu
Since 2021, EAST tokamak has been operated with full tungsten divertors. Tungsten accumulation has been frequently observed in NBI-heated H-mode discharges, resulting in the degradation of plasma confinement performance. Control of tungsten impurity is thus critical for the maintenance of high-confinement plasmas. In this work the impact of the n = 1 resonant magnetic perturbation (RMP) on the behavior of intrinsic low- and high-Z impurities in EAST H-mode discharges are experimentally studied, utilizing high-performance extreme ultraviolet spectroscopic diagnostics. In the dedicated discharge, ELM mitigation, ELM suppression, H-L back transition, RMP penetration occurs in succession with increasing RMP current (IRMP). When IRMP is below the threshold for H-L back transition, IRMP_H-L = 2.29 kA, increasing influx of C2+ and C3+ ions and decreasing influx of C4+ and C5+ ions are observed simutaneously with enhancement of the RMP field. This opposite time behavior in the influx of C4+ and C3+ ion is then observed to be magnified during the RMP penetration phase. It indicates a impurity screening layer formed between the locations where C4+ and C3+ ions distribute during RMP application based on our previous analysis (W.M. Zhang et al 2024 Nucl. Fusion 64 086004). A large step of increase in C4+ influx after H-L back transition indicates C4+ ion mainly located at bottom of pedestal. A higher RMP coil currents threshold capable of impurity screening is found for high-Z impurity ions of Cu25+, Mo30+, W42+, i. e. 0.53–0.75 kA, than that for C4+ and C5+, i. e. 0.33 kA. Meanwhile, it is found that comparing to C4+ and C5+ ions the decontamination effect by this impurity screening layer is more efficient for these high-Z impurity ions in plasma core region, e.g. up to 70 % reduction in the impurity density, leading to a significant reduction of radiation power. Furthermore, the continuous reduction of core high-Z impurities level both in ELM mitigation and suppression phase proved that this impurity decontamination effect by RMP field is dominant over the impact of ELM activity to core high-Z impurities transport since tungsten is frequently observed to accumulate during original ELM-free phase. Experimental results from this work would contribute to further understanding of the underlying mechanism how the RMP field impacts the impurity transport.
自2021年以来，EAST托卡马克一直使用全钨分流器运行。在 NBI 加热 H 模式放电中经常观察到钨积累，导致等离子体约束性能下降。因此，控制钨杂质对维持高约束等离子体至关重要。在这项工作中，利用高性能极紫外光谱诊断技术，实验研究了 n = 1 共振磁扰动（RMP）对 EAST H 模式放电中内在低 Z 和高 Z 杂质行为的影响。在专用放电中，随着 RMP 电流（IRMP）的增加，ELM 减弱、ELM 抑制、H-L 回变、RMP 穿透依次发生。当 IRMP 低于 H-L 反向转换的阈值（IRMP_H-L = 2.29 kA）时，随着 RMP 场的增强，C2+ 和 C3+ 离子的流入量增加，C4+ 和 C5+ 离子的流入量减少。在 RMP 穿透阶段，C4+ 和 C3+ 离子流入量的这种相反时间行为被放大。根据我们之前的分析（W.M. Zhang et al 2024 Nucl. Fusion 64 086004），这表明在 RMP 应用期间，C4+ 和 C3+ 离子分布位置之间形成了杂质屏蔽层。H-L 反向转换后 C4+ 流入量的大幅增加表明 C4+ 离子主要位于基座底部。与 C4+ 和 C5+ 的 0.33 kA 相比，Cu25+、Mo30+、W42+ 等高 Z 杂质离子的 RMP 线圈电流阈值更高，即 0.53-0.75 kA。同时还发现，与 C4+ 和 C5+ 离子相比，该杂质屏蔽层对等离子体核心区的这些高 Z 杂质离子的去污效果更有效，例如杂质密度可降低 70%，从而显著降低辐射功率。此外，ELM 减弱阶段和抑制阶段堆芯高 Z 杂质含量的持续降低证明，RMP 场的杂质净化效果要优于 ELM 活动对堆芯高 Z 杂质传输的影响，因为在最初的无 ELM 阶段经常观察到钨的积累。这项工作的实验结果将有助于进一步了解 RMP 磁场如何影响杂质迁移的基本机制。
{"title":"Effective control of intrinsic impurities using n = 1 resonant magnetic perturbation (RMP) in EAST H-mode plasma","authors":"Wenmin Zhang , Ling Zhang , Shigeru Morita , Yunxin Cheng , Hui Sheng , Chengxi Zhou , Huihui Wang , Youwen Sun , Yuqi Chu , Ning Sun , Ailan Hu , Darío Mitnik , Yinxian Jie , Haiqing Liu","doi":"10.1016/j.nme.2024.101822","DOIUrl":"10.1016/j.nme.2024.101822","url":null,"abstract":"<div><div>Since 2021, EAST tokamak has been operated with full tungsten divertors. Tungsten accumulation has been frequently observed in NBI-heated H-mode discharges, resulting in the degradation of plasma confinement performance. Control of tungsten impurity is thus critical for the maintenance of high-confinement plasmas. In this work the impact of the <em>n</em> = 1 resonant magnetic perturbation (RMP) on the behavior of intrinsic low- and high-Z impurities in EAST H-mode discharges are experimentally studied, utilizing high-performance extreme ultraviolet spectroscopic diagnostics. In the dedicated discharge, ELM mitigation, ELM suppression, H-L back transition, RMP penetration occurs in succession with increasing RMP current (<em>I</em><sub>RMP</sub>). When <em>I</em><sub>RMP</sub> is below the threshold for H-L back transition, <em>I</em><sub>RMP_H-L</sub> = 2.29 kA, increasing influx of C<sup>2+</sup> and C<sup>3+</sup> ions and decreasing influx of C<sup>4+</sup> and C<sup>5+</sup> ions are observed simutaneously with enhancement of the RMP field. This opposite time behavior in the influx of C<sup>4+</sup> and C<sup>3+</sup> ion is then observed to be magnified during the RMP penetration phase. It indicates a impurity screening layer formed between the locations where C<sup>4+</sup> and C<sup>3+</sup> ions distribute during RMP application based on our previous analysis (W.M. Zhang <em>et al</em> 2024 Nucl. Fusion 64 086004). A large step of increase in C<sup>4+</sup> influx after H-L back transition indicates C<sup>4+</sup> ion mainly located at bottom of pedestal. A higher RMP coil currents threshold capable of impurity screening is found for high-Z impurity ions of Cu<sup>25+</sup>, Mo<sup>30+</sup>, W<sup>42+</sup>, i. e. 0.53–0.75 kA, than that for C<sup>4+</sup> and C<sup>5+</sup>, i. e. 0.33 kA. Meanwhile, it is found that comparing to C<sup>4+</sup> and C<sup>5+</sup> ions the decontamination effect by this impurity screening layer is more efficient for these high-Z impurity ions in plasma core region, e.g. up to 70 % reduction in the impurity density, leading to a significant reduction of radiation power. Furthermore, the continuous reduction of core high-Z impurities level both in ELM mitigation and suppression phase proved that this impurity decontamination effect by RMP field is dominant over the impact of ELM activity to core high-Z impurities transport since tungsten is frequently observed to accumulate during original ELM-free phase. Experimental results from this work would contribute to further understanding of the underlying mechanism how the RMP field impacts the impurity transport.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101822"},"PeriodicalIF":2.3,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142707201","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 : 2024-11-22DOI: 10.1016/j.nme.2024.101818
N. Varadarajan , H. Bufferand , G. Ciraolo , L. Cappelli , P. Tamain , N. Rivals , S. Sureshkumar , V. Quadri , L. de Gianni
The W impurity species was introduced in the multi-component edge fluid code SOLEDGE3X, in its 2D transport version. To do so, several physical ingredients have been implemented. Firstly, W gross erosion is calculated using the Garcia-Rosales formula, and W re-deposition is estimated with either a Groth–Tskhakaya formulation or a Neural Network trained on re-deposition data from a kinetic model. This fluid approach for W transport is compared with kinetic results obtained with ERO2.0 in edge and SOL domains. Moreover, as W radiates mainly in the core region, which was thus far not resolved in SOLEDGE3X, a 1D model has been implemented for the core plasma region assuming poloidal symmetry. The impact of Oxygen impurities on W content is also addressed (including in particular the impact of W oxide layers on the PFCs). A few applications of the model implementing a Deuterium plasma with Oxygen and Tungsten, taking into account the sputtering of Tungsten by Oxygen are shown.
在多组分边缘流体代码 SOLEDGE3X 的二维传输版本中引入了 W 杂质。为此，采用了几种物理成分。首先，使用 Garcia-Rosales 公式计算 W 的总侵蚀量，然后使用 Groth-Tskhakaya 公式或根据动力学模型的再沉积数据训练的神经网络估算 W 的再沉积量。这种流体 W 传输方法与ERO2.0 在边缘域和 SOL 域获得的动力学结果进行了比较。此外，由于 W 主要在核心区域辐射，而 SOLEDGE3X 迄今为止尚未解决这一问题，因此针对核心等离子体区域实施了一个假定极对称的一维模型。此外，还讨论了氧杂质对 W 含量的影响（尤其包括 W 氧化层对全氟化碳的影响）。图中显示了该模型在氘等离子体与氧和钨之间的一些应用，并考虑了氧对钨的溅射。
{"title":"SOLEDGE3X integrated core-edge modelling of tungsten sources, migration, and radiation in WEST plasmas","authors":"N. Varadarajan , H. Bufferand , G. Ciraolo , L. Cappelli , P. Tamain , N. Rivals , S. Sureshkumar , V. Quadri , L. de Gianni","doi":"10.1016/j.nme.2024.101818","DOIUrl":"10.1016/j.nme.2024.101818","url":null,"abstract":"<div><div>The W impurity species was introduced in the multi-component edge fluid code SOLEDGE3X, in its 2D transport version. To do so, several physical ingredients have been implemented. Firstly, W gross erosion is calculated using the Garcia-Rosales formula, and W re-deposition is estimated with either a Groth–Tskhakaya formulation or a Neural Network trained on re-deposition data from a kinetic model. This fluid approach for W transport is compared with kinetic results obtained with ERO2.0 in edge and SOL domains. Moreover, as W radiates mainly in the core region, which was thus far not resolved in SOLEDGE3X, a 1D model has been implemented for the core plasma region assuming poloidal symmetry. The impact of Oxygen impurities on W content is also addressed (including in particular the impact of W oxide layers on the PFCs). A few applications of the model implementing a Deuterium plasma with Oxygen and Tungsten, taking into account the sputtering of Tungsten by Oxygen are shown.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101818"},"PeriodicalIF":2.3,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142723783","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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