Pub Date : 2024-10-28DOI: 10.1016/j.nme.2024.101782
Andreas Holm , Robert S. Wilcox , Jonathan H. Yu , Thomas D. Rognlien , Marvin E. Rensink , Filippo Scotti , Roberto Maurizio , Steve L. Allen , Wilkie Choi , Anothony W. Leonard , Morgan W. Shafer , Mathias Groth , Adam McLean
Edge-plasma simulations of a baffled, long-legged divertor in DIII-D, performed using the multi-fluid code UEDGE, indicate that the position of the detachment front is constrained to the location of the pump duct along the low-field side (LFS) baffle. Simulations including magnetic and drifts were performed for 12.5 MW deuterium plasmas including intrinsic carbon and seeded neon to assess the optimal location of the LFS divertor pump to create a stable detachment front between the target and the X-point. The radiation front position in the simulations, taken to be indicative of the detachment front, can be controlled between the pump and X-point in the favorable magnetic field direction for H-mode access by moving the pump duct location upstream of the target along the LFS baffle. In the unfavorable magnetic field direction, the radial drift flows are directed towards the pumping surface, efficiently removing the injected deuterium gas and limiting the sensitivity of the radiation front location to the gas injection rate. The role of pumping rate and drift direction on the pumping efficiency are also found to affect the divertor plasma conditions and detachment front location in UEDGE simulations.
使用多流体代码 UEDGE 对 DIII-D 中的障板长脚分流器进行的边缘等离子体模拟表明，脱离前沿的位置受限于沿低场侧（LFS）障板的泵管位置。对包括本征碳和种子氖在内的 12.5 兆瓦氘等离子体进行了包括磁漂移和 E×B 漂移在内的模拟，以评估 LFS 分流泵的最佳位置，从而在目标和 X 点之间形成稳定的脱离前沿。模拟中的辐射前沿位置被认为是脱离前沿的指示位置，可通过沿 LFS 挡板将泵管位置移至目标上游来控制泵和 X 点之间在 H 模式进入的有利磁场方向上的位置。在不利磁场方向，径向 Eθ×B 漂移流向泵表面，有效清除注入的氘气体，限制辐射锋位置对气体注入率的敏感性。在 UEDGE 模拟中还发现，抽气速率和漂移方向对抽气效率的作用也会影响分流器等离子体条件和脱离前沿位置。
{"title":"Modeling a divertor with mid-leg pumping for high-power H-mode scenarios in DIII-D considering E × B drift flows","authors":"Andreas Holm , Robert S. Wilcox , Jonathan H. Yu , Thomas D. Rognlien , Marvin E. Rensink , Filippo Scotti , Roberto Maurizio , Steve L. Allen , Wilkie Choi , Anothony W. Leonard , Morgan W. Shafer , Mathias Groth , Adam McLean","doi":"10.1016/j.nme.2024.101782","DOIUrl":"10.1016/j.nme.2024.101782","url":null,"abstract":"<div><div>Edge-plasma simulations of a baffled, long-legged divertor in DIII-D, performed using the multi-fluid code UEDGE, indicate that the position of the detachment front is constrained to the location of the pump duct along the low-field side (LFS) baffle. Simulations including magnetic and <span><math><mrow><mi>E</mi><mo>×</mo><mi>B</mi></mrow></math></span> drifts were performed for 12.5 MW deuterium plasmas including intrinsic carbon and seeded neon to assess the optimal location of the LFS divertor pump to create a stable detachment front between the target and the X-point. The radiation front position in the simulations, taken to be indicative of the detachment front, can be controlled between the pump and X-point in the favorable magnetic field direction for H-mode access by moving the pump duct location upstream of the target along the LFS baffle. In the unfavorable magnetic field direction, the radial <span><math><mrow><msub><mrow><mi>E</mi></mrow><mrow><mi>θ</mi></mrow></msub><mo>×</mo><mi>B</mi></mrow></math></span> drift flows are directed towards the pumping surface, efficiently removing the injected deuterium gas and limiting the sensitivity of the radiation front location to the gas injection rate. The role of pumping rate and drift direction on the pumping efficiency are also found to affect the divertor plasma conditions and detachment front location in UEDGE simulations.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101782"},"PeriodicalIF":2.3,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142586381","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 : 2024-10-26DOI: 10.1016/j.nme.2024.101783
J. Gerardin , Y. Corre , C. Desgranges , M. Diez , L. Dubus , M. Firdaouss , J. Gaspar , A. Grosjean , C. Guillemaut , C. Hernandez , A. Huart , H. Roche , S. Vives , WEST team
After 3 h of accumulated time from repeated plasma shots during C7 campaign performed in 2023, a deposited layer appeared on the ITER-grade W-monoblock of the lower divertor of WEST, mostly on the high field side. The growth of the deposit was observed during the campaign using infrared cameras, showing a large increase of the area covered by the deposit (x4) in the last two hours of cumulated plasma time. The deposit becomes problematic for the operation as it generates flakes which provoke radiative collapse when entering the plasma. A cleaning of the lower divertor is mandatory. A first cleaning was done using adhesive tape to remove all weakly adhered parts of the deposit. This method was chosen because it was easy to implement and did not generate dusts inside the tokamak. The cleaning enables partial removal of the more lightly adhered deposits but a large fraction remains stuck on the monoblock. A second cleaning was tried during 2024 operation by using the plasma as cleaner. A scenario was developed to put the inner strike line directly on the deposit to heat it and try to remove it by thermal stress. The deposit reaches temperature up to 1560°C but was not removed. The impurities generated were higher than normal operation and decreased during the cleaning session (−50% of light impurities observed at the end of the cleaning discharge session), showing an effect of cleaning by removing impurities from the deposit.
{"title":"Evolution and cleaning of the deposit layers on the lower divertor of WEST fully equipped with ITER grade components","authors":"J. Gerardin , Y. Corre , C. Desgranges , M. Diez , L. Dubus , M. Firdaouss , J. Gaspar , A. Grosjean , C. Guillemaut , C. Hernandez , A. Huart , H. Roche , S. Vives , WEST team","doi":"10.1016/j.nme.2024.101783","DOIUrl":"10.1016/j.nme.2024.101783","url":null,"abstract":"<div><div>After 3 h of accumulated time from repeated plasma shots during C7 campaign performed in 2023, a deposited layer appeared on the ITER-grade W-monoblock of the lower divertor of WEST, mostly on the high field side. The growth of the deposit was observed during the campaign using infrared cameras, showing a large increase of the area covered by the deposit (x4) in the last two hours of cumulated plasma time. The deposit becomes problematic for the operation as it generates flakes which provoke radiative collapse when entering the plasma. A cleaning of the lower divertor is mandatory. A first cleaning was done using adhesive tape to remove all weakly adhered parts of the deposit. This method was chosen because it was easy to implement and did not generate dusts inside the tokamak. The cleaning enables partial removal of the more lightly adhered deposits but a large fraction remains stuck on the monoblock. A second cleaning was tried during 2024 operation by using the plasma as cleaner. A scenario was developed to put the inner strike line directly on the deposit to heat it and try to remove it by thermal stress. The deposit reaches temperature up to 1560°C but was not removed. The impurities generated were higher than normal operation and decreased during the cleaning session (−50% of light impurities observed at the end of the cleaning discharge session), showing an effect of cleaning by removing impurities from the deposit.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101783"},"PeriodicalIF":2.3,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142573367","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 : 2024-10-24DOI: 10.1016/j.nme.2024.101787
Jannik Tweer , Robin Day , Thomas Derra , Daniel Dorow-Gerspach , Stefan Gräfe , Marcin Rasinski , Marius Wirtz , Christian Linsmeier , Thomas Bergs , Ghaleb Natour
Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading [1]. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energy-deposition (DED). The wire–based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs.
It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 × 12 × 5 mm3, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW m−2 (FHF = 12 MW s0.5 m−2) at a base temperature of Tbase = 700 °C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method (FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.
{"title":"Repair of heat load damaged plasma–facing material using the wire-based laser metal deposition process","authors":"Jannik Tweer , Robin Day , Thomas Derra , Daniel Dorow-Gerspach , Stefan Gräfe , Marcin Rasinski , Marius Wirtz , Christian Linsmeier , Thomas Bergs , Ghaleb Natour","doi":"10.1016/j.nme.2024.101787","DOIUrl":"10.1016/j.nme.2024.101787","url":null,"abstract":"<div><div>Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading <span><span>[1]</span></span>. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energy-deposition (DED). The wire–based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs.</div><div>It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 × 12 × 5<!--> <!-->mm<sup>3</sup>, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 10<sup>5</sup> thermal shocks of 0.5 ms duration and an intensity of <em>L</em><sub><em>abs</em></sub> = 0.55 GW m<sup>−2</sup> (<em>F</em><sub><em>HF</em></sub> = 12 MW s<sup>0</sup><sup>.</sup><sup>5</sup> m<sup>−2</sup>) at a base temperature of <em>T</em><sub><em>base</em></sub> = 700 °C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method<!--> <!-->(FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101787"},"PeriodicalIF":2.3,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552429","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 : 2024-10-24DOI: 10.1016/j.nme.2024.101788
Y. Anquetin , J. Gaspar , Y. Corre , JL. Gardarein , J. Gerardin , P. Malard , F. Rigollet , Q. Tichit , E. Tsitrone , the WEST team , the EUROfusion Tokamak Exploitation team
The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m2 and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (Pdiv) with a threshold dependant of the plasma current (IP).
估算托卡马克装置中的热通量密度分布图是一个非常重要的研究课题，不仅可以用于边缘等离子体物理目的，还能确保机器的安全。在径向方向上，热通量呈现指数衰减，分布在面向等离子体的部件中的热传感器可以捕捉到这种衰减。基于光纤布拉格光栅技术的径向分布式热传感器已嵌入 WEST 下部分流器，用于研究等离子体运行期间的热通量沉积曲线。嵌入式测量结果与三维有限元模型之间的比较显示，在衰减长度约为 30 至 50 毫米的较宽热通量顶部，存在一个较小的衰减长度（5 - 10 毫米）。为了从稳定状态下的嵌入式温度测量值预测描述沉积热通量的不同参数值，我们开发了一种使用神经网络的工具。使用该工具对在 WEST 进行的 250 多个附带分流器配置的 L 模式脉冲实验数据库进行了分析，结果表明沉积热通量跨度很大，最大热通量从 1 到 9 兆瓦/平方米不等，衰减长度从 5 到 50 毫米不等。将预测的热通量参数值与宏观等离子体参数进行比较后发现，随着分流器功率负荷（Pdiv）的增加，会出现窄分量，其阈值取决于等离子体电流（IP）。
{"title":"Identification of a double decay length (λqt) heat flux deposition shape with embedded thermal measurement and neural network","authors":"Y. Anquetin , J. Gaspar , Y. Corre , JL. Gardarein , J. Gerardin , P. Malard , F. Rigollet , Q. Tichit , E. Tsitrone , the WEST team , the EUROfusion Tokamak Exploitation team","doi":"10.1016/j.nme.2024.101788","DOIUrl":"10.1016/j.nme.2024.101788","url":null,"abstract":"<div><div>The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m<sup>2</sup> and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (P<sub>div</sub>) with a threshold dependant of the plasma current (I<sub>P</sub>).</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101788"},"PeriodicalIF":2.3,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552428","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 : 2024-10-23DOI: 10.1016/j.nme.2024.101780
S. Sureshkumar , N. Rivals , P. Tamain , X. Bonnin , R. Pitts , Y. Marandet , G. Ciraolo , H. Bufferand , G. Falchetto , N. Fedorczak , V. Quadri , M. Raghunathan , F. Schwander , E. Serre , R. Düll , N. Varadarajan
Boundary plasma simulations are essential to estimate expected divertor and first wall (FW) heat and particle loads on ITER during burning plasma operation. A key missing feature of existing SOLPS simulations (Pitts et al., 2019) is the absence of a plasma solution out to the main chamber walls, essential to self-consistently estimate the gross sputtering of wall material. Here, SOLEDGE3X is applied for the first time to obtain up-to-the wall burning plasma solutions of the ITER boundary plasma at the nominal = 100 MW of the main SOLPS database simulations, including He ash, Ne seeding but without fluid drifts. Compared with the most recent SOLPS-ITER simulations, our simulations show differences in the exact impurity distribution, but the key results for divertor and wall heat flux remain consistent. In the context of the ITER re-baselining exercise (Pitts, 2024), in which the Be FW armour is proposed to be exchanged for tungsten (W), estimates of W wall sources are key to the assessment of likely core contamination and hence impact on fusion gain. We compare the W gross erosion rates due to the different species excluding W self-sputtering. For the cases simulated spanning 0.27%–0.47% separatrix-averaged Ne concentration and D fuelling, remains the largest contributor to the sputtering flux with the largest source being the outer divertor and baffle. The species-wise contribution to W sputtering changes with fuelling with sputtering due to lower Ne charge states being significant at low D fuelling. In general, the gross W sputtering source is found to decrease with increase in D fuelling and increase with increased Ne seeding.
边界等离子体模拟对于估算燃烧等离子体运行期间热核聚变实验堆上的预期分流器和第一壁（FW）热负荷和粒子负荷至关重要。现有 SOLPS 仿真（Pitts 等人，2019 年）缺少的一个关键特征是缺少主室壁外的等离子体解决方案，而这对于自洽地估算壁材料的总溅射量至关重要。在这里，SOLEDGE3X 被首次应用于在主要 SOLPS 数据库模拟的标称 PSOL = 100 MW 条件下获得热核实验堆边界等离子体的直至壁面燃烧的等离子体解，包括 He ash、Ne seeding，但不包括流体漂移。与最新的 SOLPS-ITER 模拟相比，我们的模拟在杂质的精确分布方面存在差异，但岔流和壁面热通量的关键结果保持一致。在国际热核聚变实验堆（ITER）重新换衬底工作（Pitts，2024 年）的背景下，建议将铍 FW 盔甲换成钨（W），对 W 壁源的估计是评估可能的堆芯污染并进而影响核聚变增益的关键。我们比较了不同种类（不包括 W 自溅射）造成的 W 总侵蚀率。在跨越 0.27%-0.47% 分离矩阵平均 Ne 浓度和 7.5×1022s-1-1.95×1023s-1 D 燃料的模拟情况下，Ne8+ 仍然是溅射通量的最大贡献者，而最大的来源是外部分流器和挡板。对 W 溅射的物种贡献随燃料量的变化而变化，在低 D 燃料量下，较低的 Ne 电荷态对溅射的贡献很大。一般来说，W 溅射源总量会随着 D 注入量的增加而减少，并随着 Ne 注入量的增加而增加。
{"title":"First SOLEDGE3X-EIRENE simulations of the ITER Neon seeded burning plasma boundary up to the first wall","authors":"S. Sureshkumar , N. Rivals , P. Tamain , X. Bonnin , R. Pitts , Y. Marandet , G. Ciraolo , H. Bufferand , G. Falchetto , N. Fedorczak , V. Quadri , M. Raghunathan , F. Schwander , E. Serre , R. Düll , N. Varadarajan","doi":"10.1016/j.nme.2024.101780","DOIUrl":"10.1016/j.nme.2024.101780","url":null,"abstract":"<div><div>Boundary plasma simulations are essential to estimate expected divertor and first wall (FW) heat and particle loads on ITER during burning plasma operation. A key missing feature of existing SOLPS simulations (Pitts et al., 2019) is the absence of a plasma solution out to the main chamber walls, essential to self-consistently estimate the gross sputtering of wall material. Here, SOLEDGE3X is applied for the first time to obtain up-to-the wall burning plasma solutions of the ITER boundary plasma at the nominal <span><math><msub><mrow><mi>P</mi></mrow><mrow><mi>S</mi><mi>O</mi><mi>L</mi></mrow></msub></math></span> = 100 MW of the main SOLPS database simulations, including He ash, Ne seeding but without fluid drifts. Compared with the most recent SOLPS-ITER simulations, our simulations show differences in the exact impurity distribution, but the key results for divertor and wall heat flux remain consistent. In the context of the ITER re-baselining exercise (Pitts, 2024), in which the Be FW armour is proposed to be exchanged for tungsten (W), estimates of W wall sources are key to the assessment of likely core contamination and hence impact on fusion gain. We compare the W gross erosion rates due to the different species excluding W self-sputtering. For the cases simulated spanning 0.27%–0.47% separatrix-averaged Ne concentration and <span><math><mrow><mn>7</mn><mo>.</mo><mn>5</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>22</mn></mrow></msup><mspace></mspace><msup><mrow><mi>s</mi></mrow><mrow><mo>−</mo><mn>1</mn></mrow></msup><mo>−</mo><mn>1</mn><mo>.</mo><mn>95</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>23</mn></mrow></msup><mspace></mspace><msup><mrow><mi>s</mi></mrow><mrow><mo>−</mo><mn>1</mn></mrow></msup></mrow></math></span> D fuelling, <span><math><msup><mrow><mtext>Ne</mtext></mrow><mrow><mn>8</mn><mo>+</mo></mrow></msup></math></span> remains the largest contributor to the sputtering flux with the largest source being the outer divertor and baffle. The species-wise contribution to W sputtering changes with fuelling with sputtering due to lower Ne charge states being significant at low D fuelling. In general, the gross W sputtering source is found to decrease with increase in D fuelling and increase with increased Ne seeding.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101780"},"PeriodicalIF":2.3,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142525759","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 : 2024-10-23DOI: 10.1016/j.nme.2024.101779
J. Lovell , S.S. Henderson , J.M. Stobbs , A. Kirk , F. Federici , B.S. Patel , P.J. Ryan , J.R. Harrison , B.A. Lomanowski , J.D. Lore , MAST Upgrade Team
Global power balance calculations in steady state H mode plasmas varying the distance between separatrices () and the divertor configuration have been performed in MAST Upgrade. As becomes more negative, more of the power crossing the separatrix goes to the lower divertor. The inner divertor receives a higher fraction of the power exhaust in a Super-X divertor plasma compared to a conventional divertor plasma at similar negative , which is a concern for high power devices employing alternative divertor configurations for power exhaust handling. Global power accounting suggests of the input power is unaccounted for with the power loss channels quantified in this work. Charge exchange and orbit losses from the NBI could account for a large fraction of unaccounted power but it is not possible to precisely determine this without further diagnostic calibration.
{"title":"Experimental investigation of steady state power balance in double null and single null H mode plasmas in MAST Upgrade","authors":"J. Lovell , S.S. Henderson , J.M. Stobbs , A. Kirk , F. Federici , B.S. Patel , P.J. Ryan , J.R. Harrison , B.A. Lomanowski , J.D. Lore , MAST Upgrade Team","doi":"10.1016/j.nme.2024.101779","DOIUrl":"10.1016/j.nme.2024.101779","url":null,"abstract":"<div><div>Global power balance calculations in steady state H mode plasmas varying the distance between separatrices (<span><math><mrow><mi>d</mi><msub><mrow><mi>r</mi></mrow><mrow><mi>s</mi><mi>e</mi><mi>p</mi></mrow></msub></mrow></math></span>) and the divertor configuration have been performed in MAST Upgrade. As <span><math><mrow><mi>d</mi><msub><mrow><mi>r</mi></mrow><mrow><mi>s</mi><mi>e</mi><mi>p</mi></mrow></msub></mrow></math></span> becomes more negative, more of the power crossing the separatrix goes to the lower divertor. The inner divertor receives a higher fraction of the power exhaust in a Super-X divertor plasma compared to a conventional divertor plasma at similar negative <span><math><mrow><mi>d</mi><msub><mrow><mi>r</mi></mrow><mrow><mi>s</mi><mi>e</mi><mi>p</mi></mrow></msub></mrow></math></span>, which is a concern for high power devices employing alternative divertor configurations for power exhaust handling. Global power accounting suggests <span><math><mrow><mo>></mo><mtext>30</mtext><mspace></mspace><mtext>%</mtext></mrow></math></span> of the input power is unaccounted for with the power loss channels quantified in this work. Charge exchange and orbit losses from the NBI could account for a large fraction of unaccounted power but it is not possible to precisely determine this without further diagnostic calibration.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101779"},"PeriodicalIF":2.3,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552427","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 : 2024-10-22DOI: 10.1016/j.nme.2024.101772
X. Zhang , C. Marsden , M. Moscheni , E.N.J. Maartensson , A. Rengle , M. Robinson , T. O’Gorman , H.F. Lowe , E. Vekshina , S. Janhunen , A. Scarabosio , P.F. Buxton , M. Sertoli , M. Romanelli , S.A.M. McNamara , T.K. Gray , N.A. Lopez , the ST40 Team
The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group follows established H-mode scalings (ranging between 2 to 8 mm) while the narrow group falls up to 10 times below these scalings (between 0.2 and 0.8 mm), being comparable to the ion total Larmor radius rather than the ion poloidal Larmor radius. The heat flux profiles of the latter group can only be described by a multi-exponential function, rather than the single exponential function convoluted with a Gaussian. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.
在 ST40（一种高磁场、低长宽比的球形托卡马克）测量到的刮除层平行热通量衰减长度被观测到分为两组。宽的一组遵循既定的 H 模式标度（范围在 2 至 8 毫米之间），而窄的一组则比这些标度低 10 倍（在 0.2 至 0.8 毫米之间），与离子的总拉莫尔半径而不是离子的极拉莫尔半径相当。后一组的热通量曲线只能用多指数函数来描述，而不是用高斯卷积的单指数函数。据观测，窄刮除层宽度的出现与磁波动被抑制有关，这表明 SOL 中的电磁湍流水平降低了。
{"title":"Experimental observations of bifurcated power decay lengths in the near Scrape-Off Layer of ST40 High Field Spherical Tokamak","authors":"X. Zhang , C. Marsden , M. Moscheni , E.N.J. Maartensson , A. Rengle , M. Robinson , T. O’Gorman , H.F. Lowe , E. Vekshina , S. Janhunen , A. Scarabosio , P.F. Buxton , M. Sertoli , M. Romanelli , S.A.M. McNamara , T.K. Gray , N.A. Lopez , the ST40 Team","doi":"10.1016/j.nme.2024.101772","DOIUrl":"10.1016/j.nme.2024.101772","url":null,"abstract":"<div><div>The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group follows established H-mode scalings (ranging between 2 to 8 mm) while the narrow group falls up to 10 times below these scalings (between 0.2 and 0.8 mm), being comparable to the ion total Larmor radius rather than the ion poloidal Larmor radius. The heat flux profiles of the latter group can only be described by a multi-exponential function, rather than the single exponential function convoluted with a Gaussian. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101772"},"PeriodicalIF":2.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142573369","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 : 2024-10-22DOI: 10.1016/j.nme.2024.101773
C. Marsden , X. Zhang , M. Moscheni , T.K. Gray , E. Vekshina , A. Rengle , A. Scarabosio , M. Sertoli , M. Romanelli , ST40 team
Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m2. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.
下一代托卡马克设备中的等离子体面组件（PFC）将在极具挑战性的环境中运行，其热负荷预计将超过 10 MW/m2。这些热负荷的大小取决于排出热量的通道--"刮除层"（SOL）的宽度，并可通过通道上热通量衰减的电子折叠长度尺度来描述。随着托卡马克向反应堆相关条件发展，预计这一通道将逐渐变窄。要预测 PFC 在未来设备中必须处理的热负荷，就必须了解设置 SOL 热通量宽度所涉及的过程。在高场球形托卡马克 ST40 上使用新投入使用的红外热成像系统对 SOL 宽度进行了测量。ST40 具有高轴环形磁场（≥1.5 T），在研究环形磁场对球形托卡马克热通量宽度的影响方面具有得天独厚的优势，同时还能将球形托卡马克中的 SOL 宽度测量扩展到更高的极性磁场（≥0.3 T）。由于 ST40 上的分流器轴对称程度较低，因此有必要对整个分流器的热通量进行径向测量。这些径向剖面图与等离子体的磁性几何形状相结合，推断出 SOL 的宽度，并观察到热通量的艾希曲线和双指数曲线。在分流器的部分环形区域观察到热通量的减少，同时在瓦片边缘的局部区域观察到热负荷的增加。为了进一步了解所观测到的热通量模式，需要在今后的工作中确定瓦片错位的影响以及重建的岔道磁几何形状的不确定性，还需要对不均匀鞘电场可能发挥的作用进行更多的研究。
{"title":"Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry","authors":"C. Marsden , X. Zhang , M. Moscheni , T.K. Gray , E. Vekshina , A. Rengle , A. Scarabosio , M. Sertoli , M. Romanelli , ST40 team","doi":"10.1016/j.nme.2024.101773","DOIUrl":"10.1016/j.nme.2024.101773","url":null,"abstract":"<div><div>Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m<sup>2</sup>. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (<span><math><mo>≥</mo></math></span>1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (<span><math><mo>≥</mo></math></span>0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101773"},"PeriodicalIF":2.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142561033","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 : 2024-10-22DOI: 10.1016/j.nme.2024.101786
Martina Molinari , Eugenio Lo Piccolo , Raffaele Torella , Matteo D’Onorio , Nicholas Terranova , Gianfranco Caruso
This paper presents experimental findings on the behavior of iron-based alloys in environmental conditions typical of nuclear fusion technology, specifically focusing on material degradation, which is a critical aspect for the water cooling system of EU DEMO breeding blankets. The experimental campaign investigates potassium hydroxide’s role as an alkalizing agent, testing various concentrations to assess its impact on corrosion resistance. Additionally, it examines how oxygen levels affect localized corrosion development, which is crucial for mitigating corrosion risks in fusion applications. Seven 1000-hour tests were conducted to determine optimal conditions for corrosion reduction. Findings include identifying an oxygen concentration threshold to prevent piping cracking on EUROFER97 specimens.
{"title":"Experimental insights on iron-based alloys corrosion in water cooled loops","authors":"Martina Molinari , Eugenio Lo Piccolo , Raffaele Torella , Matteo D’Onorio , Nicholas Terranova , Gianfranco Caruso","doi":"10.1016/j.nme.2024.101786","DOIUrl":"10.1016/j.nme.2024.101786","url":null,"abstract":"<div><div>This paper presents experimental findings on the behavior of iron-based alloys in environmental conditions typical of nuclear fusion technology, specifically focusing on material degradation, which is a critical aspect for the water cooling system of EU DEMO breeding blankets. The experimental campaign investigates potassium hydroxide’s role as an alkalizing agent, testing various concentrations to assess its impact on corrosion resistance. Additionally, it examines how oxygen levels affect localized corrosion development, which is crucial for mitigating corrosion risks in fusion applications. Seven 1000-hour tests were conducted to determine optimal conditions for corrosion reduction. Findings include identifying an oxygen concentration threshold to prevent piping cracking on EUROFER97 specimens.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101786"},"PeriodicalIF":2.3,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142525758","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 : 2024-10-21DOI: 10.1016/j.nme.2024.101785
Zhenhua Hu , Xue Bai , Huace Wu , Ran Hai , Fang Ding , Muhammad Imran , Cong Li , Hongbin Ding , Guang-Nan Luo
In this work, the characterization of impurities deposited on the Plasma-Facing Components (PFCs) of the EAST tokamak, — including the tungsten upper and lower divertors (UD, LD), the molybdenum first wall on the high-field side (HFS), and the graphite main guard limiter (ML) — is reported using a portable Laser-Induced Breakdown Spectroscopy (LIBS) device. The LIBS analysis revealed that the primary impurity elements deposited on the PFCs are Cu, W, Fe, Li, Mo, and Ca. Notably, significant amounts of Cu, W, Fe, Li, and Mo impurities were observed on the ML surface. On both the upper and lower divertors (UD, LD) surfaces, trace amounts of Cu and Fe impurities were detected. The results on HFS showed a low level of Cu, W, and Fe impurities on its surface. Additionally, the Calibration-Free LIBS (CF-LIBS) method was implemented to determine the relative content of impurities on the PFCs. The quantitative results of the deposited impurities on PFCs further provide a more detailed interpretation of the variations in impurity deposition. The successful in situ assessment of impurities deposited on the EAST tokamak using a portable LIBS device demonstrates the potential for integrating the LIBS system into a remote handling system. These results offer valuable insights into the dynamics of impurities in controlled fusion devices, which can contribute to enhancing impurity control methods and improving device performance.
在这项工作中，使用便携式激光诱导击穿光谱（LIBS）设备报告了沉积在 EAST 托卡马克面向等离子体的部件（PFC）上的杂质特征，包括钨上下分流器（UD、LD）、高场侧钼第一壁（HFS）和石墨主防护限制器（ML）。LIBS 分析显示，沉积在 PFC 上的主要杂质元素是铜、钨、铁、锂、钼和钙。值得注意的是，在 ML 表面观察到了大量的 Cu、W、Fe、Li 和 Mo 杂质。在上分流器和下分流器（UD、LD）表面都检测到了微量的铜和铁杂质。HFS 的结果显示其表面的铜、钨和铁杂质含量较低。此外，还采用了无校准 LIBS（CF-LIBS）方法来确定 PFC 上杂质的相对含量。PFC 上沉积杂质的定量结果进一步详细解释了杂质沉积的变化。使用便携式 LIBS 设备成功地对 EAST 托卡马克上沉积的杂质进行了现场评估，证明了将 LIBS 系统集成到远程处理系统中的潜力。这些结果为了解受控核聚变装置中杂质的动态提供了宝贵的见解，有助于改进杂质控制方法和提高装置性能。
{"title":"Quantitative analysis of impurities deposited on the Plasma-Facing Components of EAST tokamak using a portable LIBS device","authors":"Zhenhua Hu , Xue Bai , Huace Wu , Ran Hai , Fang Ding , Muhammad Imran , Cong Li , Hongbin Ding , Guang-Nan Luo","doi":"10.1016/j.nme.2024.101785","DOIUrl":"10.1016/j.nme.2024.101785","url":null,"abstract":"<div><div>In this work, the characterization of impurities deposited on the Plasma-Facing Components (PFCs) of the EAST tokamak, — including the tungsten upper and lower divertors (UD, LD), the molybdenum first wall on the high-field side (HFS), and the graphite main guard limiter (ML) — is reported using a portable Laser-Induced Breakdown Spectroscopy (LIBS) device. The LIBS analysis revealed that the primary impurity elements deposited on the PFCs are Cu, W, Fe, Li, Mo, and Ca. Notably, significant amounts of Cu, W, Fe, Li, and Mo impurities were observed on the ML surface. On both the upper and lower divertors (UD, LD) surfaces, trace amounts of Cu and Fe impurities were detected. The results on HFS showed a low level of Cu, W, and Fe impurities on its surface. Additionally, the Calibration-Free LIBS (CF-LIBS) method was implemented to determine the relative content of impurities on the PFCs. The quantitative results of the deposited impurities on PFCs further provide a more detailed interpretation of the variations in impurity deposition. The successful <em>in situ</em> assessment of impurities deposited on the EAST tokamak using a portable LIBS device demonstrates the potential for integrating the LIBS system into a remote handling system. These results offer valuable insights into the dynamics of impurities in controlled fusion devices, which can contribute to enhancing impurity control methods and improving device performance.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101785"},"PeriodicalIF":2.3,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552425","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}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号