Pub Date : 2024-03-02DOI: 10.1088/1741-4326/ad2f4e
Tullio Barbui, Luis Delgado-Aparicio, Brentley Stratton, O. Chellaï, R. Dumont, Kenneth W Hill, N. Pablant, Yves Peysson
A new method to obtain the mean energy of fast electron losses in fusion plasmas using a versatile multi-energy hard x-ray detector is presented. The method is based on measuring the thick-target emission of tungsten in the divertor region produced by fast electron losses interacting with the target and modeling the tungsten spectra by a Monte Carlo code which simulates the interaction between a beam of electrons and a solid target. The mean energy of the fast electron losses is determined through the comparison between the experimental and synthetic emission. The results show that fast electron losses during lower hybrid current drive discharges at WEST have a mean energy of 90-140 keV and represent only 2% of the total heat flux at the target. Additionally, anisotropic hard x-ray emission has been detected for the first time at the WEST core and edge plasma, with opposite directions. It is due to the forward-peak emission of two distinctive populations of fast electrons: co-current fast electrons in the core and counter-current fast electron losses at the inner strike point. In view of future experiments like ITER where electron cyclotron current drive will generate a fast electron population, this technique could serve as a real-time monitor of fast electron losses and eventually feed an actuator on the current drive generation.
介绍了一种利用多功能硬X射线探测器获得聚变等离子体中快速电子损耗平均能量的新方法。该方法基于测量由快速电子损耗与靶相互作用产生的钨在分流区的厚靶发射，并通过蒙特卡洛代码模拟电子束与固体靶之间的相互作用来建立钨光谱模型。通过比较实验发射和合成发射，确定了快速电子损耗的平均能量。结果表明，在 WEST 的较低混合电流驱动放电过程中，快速电子损耗的平均能量为 90-140 keV，仅占靶上总热流量的 2%。此外，在 WEST 核心和边缘等离子体中首次探测到了方向相反的各向异性硬 X 射线发射。这是由于两种不同的快速电子群向前峰发射所致：内核的同流快速电子和内撞击点的逆流快速电子损耗。鉴于未来的实验（如热核实验堆）中电子回旋电流驱动将产生快速电子群，该技术可作为快速电子损耗的实时监测器，并最终为电流驱动产生的致动器提供信息。
{"title":"Determination of the mean energy of fast electron losses and anisotropies through thick-target emission on WEST","authors":"Tullio Barbui, Luis Delgado-Aparicio, Brentley Stratton, O. Chellaï, R. Dumont, Kenneth W Hill, N. Pablant, Yves Peysson","doi":"10.1088/1741-4326/ad2f4e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2f4e","url":null,"abstract":"\u0000 A new method to obtain the mean energy of fast electron losses in fusion plasmas using a versatile multi-energy hard x-ray detector is presented. The method is based on measuring the thick-target emission of tungsten in the divertor region produced by fast electron losses interacting with the target and modeling the tungsten spectra by a Monte Carlo code which simulates the interaction between a beam of electrons and a solid target. The mean energy of the fast electron losses is determined through the comparison between the experimental and synthetic emission. The results show that fast electron losses during lower hybrid current drive discharges at WEST have a mean energy of 90-140 keV and represent only 2% of the total heat flux at the target. Additionally, anisotropic hard x-ray emission has been detected for the first time at the WEST core and edge plasma, with opposite directions. It is due to the forward-peak emission of two distinctive populations of fast electrons: co-current fast electrons in the core and counter-current fast electron losses at the inner strike point. In view of future experiments like ITER where electron cyclotron current drive will generate a fast electron population, this technique could serve as a real-time monitor of fast electron losses and eventually feed an actuator on the current drive generation.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"30 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-03-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140081569","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-03-02DOI: 10.1088/1741-4326/ad2f4c
Long Li, Zhe Liu, Ze Chen, Chao Yin, S. Mao, X. B. Wu, Noriyasu Ohno, Minyou Ye
ZrC dispersion-strengthened W exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. In this study, surface modification of 0.5 wt.% ZrC dispersion-strengthened W (WZrC) under low energy and high fluence He plasma irradiation at high temperature was presented. Under the energy of 90 eV and fluence ranging from 6 × 1024 He∙m-2 to 2 × 1026 He∙m-2 He irradiation at 920 ℃, typical fuzz nanostructure appeared on the W matrix of WZrC. The fuzz showed comparable thickness and structure features to pure W, which indicates limited effects of the particle’s addition on resistance to high fluence He irradiation at high temperatures. Besides, the erosion behavior of particles under He plasma irradiation has been investigated, which is thought to be dominated by a sputtering process. Under the He influence of 6 × 1024 He∙m-2, only nanopores were observed in the surface region. With fluence increasing to 5 × 1025 He∙m-2, the surface became relatively uneven with larger holes and stalagmitic structures. And W accumulated on the top of stalagmitic structures due to the subthreshold sputtering under He irradiation. When fluence further increased to 2 × 1026 He∙m-2, the particles were eroded completely and covered by the extended fuzz, forming cavities. In addition, distinctive layered nanotendrils were observed above the cavities, which were characterized to be consist of inner W-riched skeletons and outer Zr-riched layers. It indicates that the layered nanotendrils should be the result of fuzz extension combined with particles sputtering and redeposition.
ZrC 分散强化 W 具有高强度/韧性、低韧性-脆性转变温度和优异的抗热震性，因此有望成为未来聚变装置的候选等离子体面材料。本研究介绍了 0.5 wt.% ZrC 分散强化 W（WZrC）在低能量、高通量 He 等离子体高温辐照下的表面改性情况。在 90 eV 的能量和 6 × 1024 He∙m-2 到 2 × 1026 He∙m-2 的通量范围内，在 920 ℃ 的等离子体辐照下，WZrC 的 W 基体上出现了典型的模糊纳米结构。毛细管的厚度和结构特征与纯 W 相当，这表明颗粒的添加对高温下抵抗高通量 He 辐照的影响有限。此外，还研究了粒子在 He 等离子体辐照下的侵蚀行为。在 6 × 1024 He∙m-2 的 He 影响下，仅在表面区域观察到纳米孔。当通量增加到 5 × 1025 He∙m-2 时，表面变得相对不平整，出现了较大的孔洞和星状结构。由于阈下溅射是在 He 的辐照下进行的，因此 W 会积聚在星状结构的顶部。当通量进一步增加到 2 × 1026 He∙m-2 时，颗粒被完全侵蚀，并被扩展的绒毛覆盖，形成空穴。此外，在空穴上方还观察到了明显的分层纳米卷须，其特征是由内部的 W-riched 骨架和外部的 Zr-riched 层组成。这表明分层纳米卷须应该是模糊延伸与颗粒溅射和再沉积相结合的结果。
{"title":"Surface modification of ZrC dispersion-strengthened W under low energy He plasma irradiation","authors":"Long Li, Zhe Liu, Ze Chen, Chao Yin, S. Mao, X. B. Wu, Noriyasu Ohno, Minyou Ye","doi":"10.1088/1741-4326/ad2f4c","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2f4c","url":null,"abstract":"\u0000 ZrC dispersion-strengthened W exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. In this study, surface modification of 0.5 wt.% ZrC dispersion-strengthened W (WZrC) under low energy and high fluence He plasma irradiation at high temperature was presented. Under the energy of 90 eV and fluence ranging from 6 × 1024 He∙m-2 to 2 × 1026 He∙m-2 He irradiation at 920 ℃, typical fuzz nanostructure appeared on the W matrix of WZrC. The fuzz showed comparable thickness and structure features to pure W, which indicates limited effects of the particle’s addition on resistance to high fluence He irradiation at high temperatures. Besides, the erosion behavior of particles under He plasma irradiation has been investigated, which is thought to be dominated by a sputtering process. Under the He influence of 6 × 1024 He∙m-2, only nanopores were observed in the surface region. With fluence increasing to 5 × 1025 He∙m-2, the surface became relatively uneven with larger holes and stalagmitic structures. And W accumulated on the top of stalagmitic structures due to the subthreshold sputtering under He irradiation. When fluence further increased to 2 × 1026 He∙m-2, the particles were eroded completely and covered by the extended fuzz, forming cavities. In addition, distinctive layered nanotendrils were observed above the cavities, which were characterized to be consist of inner W-riched skeletons and outer Zr-riched layers. It indicates that the layered nanotendrils should be the result of fuzz extension combined with particles sputtering and redeposition.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"46 9","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-03-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140082310","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-02-23DOI: 10.1088/1741-4326/ad2ca8
Qiming Hu, N. Logan, Qingquan Yu, A. Bortolon
According to recent DIII-D experiments (N.C. Logan et al 2024 Nucl.Fusion 64 014003), injecting edge localized electron cyclotron current drive (ECCD) in the counter-plasma-current (counter-Ip) direction reduces the n = 3 resonant magnetic perturbation (RMP) current threshold for ELM suppression, while co-Ip ECCD during the suppressed ELM phase causes a back transition to ELMing. This paper presents nonlinear two-fluid simulations on the ECCD manipulation of edge magnetic islands induced by RMP using the TM1 code. In the presence of a magnetic island chain at the pedestal-top, co-Ip ECCD is found to decrease the island width and restore the initially degraded pedestal pressure when its radial deposition location is close to the rational surface of the island. With a sufficiently strong co-Ip ECCD current, the RMP-driven magnetic island can be healed, and the pedestal pressure fully recovers to its initial ELMing state. On the contrary, counter-Ip ECCD is found to increase the island width and further reduce the pedestal pressure to levels significantly below the peeling-ballooning-mode limited height, leading to even stationary ELM suppression. These simulations align with the results from DIII-D experiments. However, when multiple magnetic island chains are present at the pedestal-top, the ECCD current experiences substantial broadening, and its effects on the island width and pedestal pressure become negligible. Further simulations reveal that counter-Ip ECCD enhances RMP penetration by lowering the penetration threshold, with the degree of reduction proportional to the amplitude of ECCD current. For the ~1 MW ECCD in DIII-D, the predicted decrease in the RMP penetration threshold for ELM suppression is approximately 20%, consistent with experimental observations. These simulations indicate that edge-localized ECCD can be used to either facilitate RMP ELM suppression or optimize the confinement degradation.
{"title":"Effects of edge-localized electron cyclotron current drive on edge-localized mode suppression by resonant magnetic perturbations in DIII-D","authors":"Qiming Hu, N. Logan, Qingquan Yu, A. Bortolon","doi":"10.1088/1741-4326/ad2ca8","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2ca8","url":null,"abstract":"\u0000 According to recent DIII-D experiments (N.C. Logan et al 2024 Nucl.Fusion 64 014003), injecting edge localized electron cyclotron current drive (ECCD) in the counter-plasma-current (counter-Ip) direction reduces the n = 3 resonant magnetic perturbation (RMP) current threshold for ELM suppression, while co-Ip ECCD during the suppressed ELM phase causes a back transition to ELMing. This paper presents nonlinear two-fluid simulations on the ECCD manipulation of edge magnetic islands induced by RMP using the TM1 code. In the presence of a magnetic island chain at the pedestal-top, co-Ip ECCD is found to decrease the island width and restore the initially degraded pedestal pressure when its radial deposition location is close to the rational surface of the island. With a sufficiently strong co-Ip ECCD current, the RMP-driven magnetic island can be healed, and the pedestal pressure fully recovers to its initial ELMing state. On the contrary, counter-Ip ECCD is found to increase the island width and further reduce the pedestal pressure to levels significantly below the peeling-ballooning-mode limited height, leading to even stationary ELM suppression. These simulations align with the results from DIII-D experiments. However, when multiple magnetic island chains are present at the pedestal-top, the ECCD current experiences substantial broadening, and its effects on the island width and pedestal pressure become negligible. Further simulations reveal that counter-Ip ECCD enhances RMP penetration by lowering the penetration threshold, with the degree of reduction proportional to the amplitude of ECCD current. For the ~1 MW ECCD in DIII-D, the predicted decrease in the RMP penetration threshold for ELM suppression is approximately 20%, consistent with experimental observations. These simulations indicate that edge-localized ECCD can be used to either facilitate RMP ELM suppression or optimize the confinement degradation.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"34 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-02-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139957441","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-02-19DOI: 10.1088/1741-4326/ad2abb
Markus Weiland, O. Kudlacek, Bernhard Sieglin, Roberto Bilato, U. Plank, W. Treutterer
Conventionally, neutral beam injection (NBI) in tokamaks is controlled via engineering parameters such as injection voltage and power. Recently, the high-fidelity real-time NBI code RABBIT has been coupled to the Discharge Control System (DCS) of ASDEX Upgrade. It allows to calculate the NBI fast-ion distribution and hence the properties of NBI in real-time, making it possible to control them directly. We successfully demonstrate control of driven current, ion heating and stored fast-ion energy by modifying the injected beam power. A combined ECRH and NBI controller is also successfully tested, which is able to adjust the heating mix between ECRH and NBI to match a certain desired ion heating fraction at given total power. Further experiments have been carried out towards control of the ion heat flux (i.e. ion heating plus collisional heat transfer between ions and electrons). They show good initial success, but also leave room for future improvements as the controller runs into instabilities at too high requests.
{"title":"Real-time control of NBI fast ions, current-drive and heating properties","authors":"Markus Weiland, O. Kudlacek, Bernhard Sieglin, Roberto Bilato, U. Plank, W. Treutterer","doi":"10.1088/1741-4326/ad2abb","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2abb","url":null,"abstract":"\u0000 Conventionally, neutral beam injection (NBI) in tokamaks is controlled via engineering parameters such as injection voltage and power. Recently, the high-fidelity real-time NBI code RABBIT has been coupled to the Discharge Control System (DCS) of ASDEX Upgrade. It allows to calculate the NBI fast-ion distribution and hence the properties of NBI in real-time, making it possible to control them directly. We successfully demonstrate control of driven current, ion heating and stored fast-ion energy by modifying the injected beam power. A combined ECRH and NBI controller is also successfully tested, which is able to adjust the heating mix between ECRH and NBI to match a certain desired ion heating fraction at given total power. Further experiments have been carried out towards control of the ion heat flux (i.e. ion heating plus collisional heat transfer between ions and electrons). They show good initial success, but also leave room for future improvements as the controller runs into instabilities at too high requests.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"12 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139958956","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-02-16DOI: 10.1088/1741-4326/ad2a28
P. Chiovaro, A. Quartararo, Pietro Avona, G. Bongiovì, P. A. di Maio, S. Giambrone, Ivo Moscato, E. Vallone
In water-cooled nuclear reactors, the issue of neutron-activated products transport along the Primary Heat Transfer System (PHTS) is very demanding, as it is a coupled neutronic / fluid-dynamic problem requiring a challenging balance between accuracy and reasonable computational time. This work addresses the transport of water activation products in large hydraulic circuits. Regarding the nuclear calculations, the assessment of the production rates of the radioisotope concentrations has been performed by Monte Carlo analyses adopting the MCNP5.1.6 code, while for the transportation calculations, an innovative method has been expressly developed. It foresees a one-dimensional nodalization, in a MATLAB-Simulink environment, of the hydraulic circuit considered with a Computational Fluid-Dynamic (CFD) characterization (by ANSYS CFX code) of the nodes under neutron flux, that is the components where radioisotopes are formed, and the highest gradients of concentration are present. The method was compared with one-dimensional models not supported by fluid-dynamic analysis. The results of this comparison showed that in cases involving fairly complicated geometries and radioisotopes with a small half-life, CFD analyses are necessary to achieve adequate accuracy. The procedure was applied to very large and rather complex hydraulic circuits like the divertor PHTSs of DEMO fusion reactor to obtain the concentrations of the activation products of the water constituents (16N, 17N, 19O, 14C, 41Ar) along such systems.
{"title":"Water activation products generation and transport in DEMO divertor","authors":"P. Chiovaro, A. Quartararo, Pietro Avona, G. Bongiovì, P. A. di Maio, S. Giambrone, Ivo Moscato, E. Vallone","doi":"10.1088/1741-4326/ad2a28","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2a28","url":null,"abstract":"\u0000 In water-cooled nuclear reactors, the issue of neutron-activated products transport along the Primary Heat Transfer System (PHTS) is very demanding, as it is a coupled neutronic / fluid-dynamic problem requiring a challenging balance between accuracy and reasonable computational time. This work addresses the transport of water activation products in large hydraulic circuits. Regarding the nuclear calculations, the assessment of the production rates of the radioisotope concentrations has been performed by Monte Carlo analyses adopting the MCNP5.1.6 code, while for the transportation calculations, an innovative method has been expressly developed. It foresees a one-dimensional nodalization, in a MATLAB-Simulink environment, of the hydraulic circuit considered with a Computational Fluid-Dynamic (CFD) characterization (by ANSYS CFX code) of the nodes under neutron flux, that is the components where radioisotopes are formed, and the highest gradients of concentration are present. The method was compared with one-dimensional models not supported by fluid-dynamic analysis. The results of this comparison showed that in cases involving fairly complicated geometries and radioisotopes with a small half-life, CFD analyses are necessary to achieve adequate accuracy. The procedure was applied to very large and rather complex hydraulic circuits like the divertor PHTSs of DEMO fusion reactor to obtain the concentrations of the activation products of the water constituents (16N, 17N, 19O, 14C, 41Ar) along such systems.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"9 4","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139960884","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-02-16DOI: 10.1088/1741-4326/ad2a2a
M. Carpita, O. Février, Holger Reimerdes, C. Theiler, B. P. Duval, C. Colandrea, G. Durr-Legoupil-Nicoud, D. Galassi, S. Gorno, E. Huett, Joaquim Loizu, Lorenzo Martinelli, A. Perek, Luke Simons, Guangyu Sun, E. Tonello, C. Wüthrich
The Super-X divertor (SXD) is an alternative divertor configuration leveraging total flux expansion at the outer strike point (OSP). While the extended 2-point model (2PM) predicts facilitated detachment access and control in the SXD configuration, these attractive features are not always retrieved experimentally. These discrepancies are at least partially explained by the effect of parallel flows which, when self-consistently included in the 2PM, reveal the role of total flux expansion on the pressure balance and weaken the total flux expansion effect on detachment access and control, compared to the original predictions. This new model can partially explain the discrepancies between the 2PM and experiments performed on TCV, in ohmic L-mode scenarios, which are particularly apparent when scanning the OSP major radius Rt. In core density ramps in lower single-null (SN) configuration, the impact of Rt on the CIII emission front movement in the divertor outer leg - used as a proxy for the plasma temperature in the divertor – is substantially weaker than 2PM predictions. Furthermore, in OSP radial sweeps in lower and upper SN configurations, in ohmic L-mode scenarios with a constant core density, the peak parallel particle flux density at the OSP is almost independent of Rt, while the 2PM predicts a linear dependence. Finally, analytical and numerical modelling of parallel flows in the divertor is presented. It is shown that an increase in total flux expansion can favour supersonic flows at the OSP. Parallel flows are also shown to be relevant by analysing SOLPS-ITER simulations of TCV.
{"title":"Parallel flows as a key component to interpret Super-X divertor experiments","authors":"M. Carpita, O. Février, Holger Reimerdes, C. Theiler, B. P. Duval, C. Colandrea, G. Durr-Legoupil-Nicoud, D. Galassi, S. Gorno, E. Huett, Joaquim Loizu, Lorenzo Martinelli, A. Perek, Luke Simons, Guangyu Sun, E. Tonello, C. Wüthrich","doi":"10.1088/1741-4326/ad2a2a","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2a2a","url":null,"abstract":"\u0000 The Super-X divertor (SXD) is an alternative divertor configuration leveraging total flux expansion at the outer strike point (OSP). While the extended 2-point model (2PM) predicts facilitated detachment access and control in the SXD configuration, these attractive features are not always retrieved experimentally. These discrepancies are at least partially explained by the effect of parallel flows which, when self-consistently included in the 2PM, reveal the role of total flux expansion on the pressure balance and weaken the total flux expansion effect on detachment access and control, compared to the original predictions. This new model can partially explain the discrepancies between the 2PM and experiments performed on TCV, in ohmic L-mode scenarios, which are particularly apparent when scanning the OSP major radius Rt. In core density ramps in lower single-null (SN) configuration, the impact of Rt on the CIII emission front movement in the divertor outer leg - used as a proxy for the plasma temperature in the divertor – is substantially weaker than 2PM predictions. Furthermore, in OSP radial sweeps in lower and upper SN configurations, in ohmic L-mode scenarios with a constant core density, the peak parallel particle flux density at the OSP is almost independent of Rt, while the 2PM predicts a linear dependence. Finally, analytical and numerical modelling of parallel flows in the divertor is presented. It is shown that an increase in total flux expansion can favour supersonic flows at the OSP. Parallel flows are also shown to be relevant by analysing SOLPS-ITER simulations of TCV.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"51 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139961557","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-02-16DOI: 10.1088/1741-4326/ad2a29
E. Hodille, Blanche Pavec, J. Denis, Axel Dunand, Yves Ferro, Marco Minissale, T. Angot, Christian Grisolia, R. Bisson
Rate equation modelling is performed to simulate D2 and D2+D2 + exposure of the W(110) surface with varying coverage of oxygen atoms (O) from the clean surface up to 0.75 monolayer of O. Density functional Theory (DFT) calculated energetics are used as inputs for the surface processes and desorption energies are optimized to best reproduce the thermal desorption spectrometry (TDS) experiments obtained for D2 exposure. For the clean surface, the optimized desorption energies (1.10 eV to 1.40 eV) are below the DFT ones (1.30 eV to 1.50 eV). For the O covered surface, the main desorption peak is reproduced with desorption energies of 1.1 eV and 1.0 eV for 0.50 and 0.75 monolayer of O respectively. This is slightly higher than the DFT predicted desorption energies. In order to simulate satisfactorily the total retention botained experimentally for D2+D2 + exposure, a sputtering process needs to be added to the model, describing the sputtering of adsorbed species (D atoms) by the incident D ions. The impact of the sputtering process on the shape of the TDS spectra, on the total retention and on the recycling of D from the wall is discussed. In order to better characterize the sputtering process, especially its products and yields, atomistic calculations such as molecular dynamics are suggested as a next step for this study.
利用密度泛函理论（DFT）计算的能量作为表面过程的输入，并对解吸能量进行了优化，以最好地再现 D2 暴露时获得的热解吸光谱（TDS）实验结果。对于清洁表面，优化后的解吸能量（1.10 eV 至 1.40 eV）低于 DFT 能量（1.30 eV 至 1.50 eV）。对于 O 覆盖的表面，0.50 和 0.75 单层 O 的解吸能分别为 1.1 eV 和 1.0 eV，再现了主要的解吸峰。这比 DFT 预测的解吸能量略高。为了令人满意地模拟 D2+D2 + 暴露实验中的总滞留率，需要在模型中加入溅射过程，描述入射 D 离子对吸附物种（D 原子）的溅射。本文讨论了溅射过程对 TDS 光谱形状、总保留量和 D 从壁回收的影响。为了更好地描述溅射过程，特别是其产物和产量，建议将分子动力学等原子计算作为本研究的下一步。
{"title":"Deuterium uptake, desorption and sputtering from W(110) surface covered with oxygen","authors":"E. Hodille, Blanche Pavec, J. Denis, Axel Dunand, Yves Ferro, Marco Minissale, T. Angot, Christian Grisolia, R. Bisson","doi":"10.1088/1741-4326/ad2a29","DOIUrl":"https://doi.org/10.1088/1741-4326/ad2a29","url":null,"abstract":"\u0000 Rate equation modelling is performed to simulate D2 and D2+D2\u0000 + exposure of the W(110) surface with varying coverage of oxygen atoms (O) from the clean surface up to 0.75 monolayer of O. Density functional Theory (DFT) calculated energetics are used as inputs for the surface processes and desorption energies are optimized to best reproduce the thermal desorption spectrometry (TDS) experiments obtained for D2 exposure. For the clean surface, the optimized desorption energies (1.10 eV to 1.40 eV) are below the DFT ones (1.30 eV to 1.50 eV). For the O covered surface, the main desorption peak is reproduced with desorption energies of 1.1 eV and 1.0 eV for 0.50 and 0.75 monolayer of O respectively. This is slightly higher than the DFT predicted desorption energies. In order to simulate satisfactorily the total retention botained experimentally for D2+D2\u0000 + exposure, a sputtering process needs to be added to the model, describing the sputtering of adsorbed species (D atoms) by the incident D ions. The impact of the sputtering process on the shape of the TDS spectra, on the total retention and on the recycling of D from the wall is discussed. In order to better characterize the sputtering process, especially its products and yields, atomistic calculations such as molecular dynamics are suggested as a next step for this study.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"30 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139960642","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-10DOI: 10.1088/1741-4326/ad1d11
Jae-Sun Park, Xavier Bonnin, R. Pitts, J. Lore
The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database cite{pitts2019physics}, assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favours gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in cite{park2020assessment}. However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20-40 %, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.
{"title":"Impact of gas injection location and divertor surface material on ITER fusion power operation phase divertor performance assessed with SOLPS-ITER","authors":"Jae-Sun Park, Xavier Bonnin, R. Pitts, J. Lore","doi":"10.1088/1741-4326/ad1d11","DOIUrl":"https://doi.org/10.1088/1741-4326/ad1d11","url":null,"abstract":"\u0000 The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database cite{pitts2019physics}, assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favours gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in cite{park2020assessment}. However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20-40 %, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"34 11","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139441506","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-10DOI: 10.1088/1741-4326/ad1d10
B. Stein-Lubrano, R. Sweeney, Daniele Bonfiglio, J. Lovell, Pedro Carvalho, L. Baylor, R. Granetz, Stefan Jachmich, E. Joffrin, Mengdi Kong, M. Lehnen, C. Maggi, Earl S Marmar, E. Nardon, P. Puglia, U. Sheikh, Daisuke Shiraki, S. Silburn
Precise values for radiated energy in tokamak disruption experiments are needed to validate disruption mitigation techniques for burning plasma tokamaks like ITER and SPARC. Control room analysis of radiated power (Prad) on JET assumes axisymmetry, since fitting 3D radiation structures with limited bolometry coverage is an underdetermined problem. In mitigated disruptions, radiation is toroidally asymmetric and 3D, due to fast-growing 3D MHD modes and localized impurity sources. To address this problem, Emis3D adopts a physics motivated forward modeling ("guess and check") approach, comparing experimental bolometry data to synthetic data from user-defined radiation structures. Synthetic structures are observed with the Cherab modeling framework and a best fit chosen using a reduced χ2 statistic. 2D tomographic inversion models are tested, as well as helical flux tubes and 3D MHD simulated structures from JOREK. Two nominally identical pure neon shattered pellet injection (SPI) mitigated discharges in JET are analyzed. 2D tomographic inversions with added toroidal freedom are the best fits in the thermal quench (TQ) and current quench (CQ). In the pre-TQ, 2D reconstructions are statistically the best fits, but are likely over-optimized and do not capture the 3D radiation structure seen in fast camera images. The next-best pre-TQ fits are helical structures that extend towards the high-field side, consistent with an impurity flow under the magnetic nozzle effect also observed in JOREK simulations. Whole-disruption radiated fractions of 0.98 +0.03/-0.29 and 1.01 +0.02/-0.17 are found, suggesting that the stored energy may have been fully mitigated by each SPI, although mitigation efficiencies well below ITER and SPARC requirements for high energy pulses are still within the large uncertainties. Emis3D is also used to validate JOREK SPI simulations, and confirms improvements in matching experiment from changes to impurity modeling. Time-dependent toroidal peaking factors are calculated and discussed.
{"title":"3D radiated power analysis of JET SPI discharges using the Emis3D forward modeling tool","authors":"B. Stein-Lubrano, R. Sweeney, Daniele Bonfiglio, J. Lovell, Pedro Carvalho, L. Baylor, R. Granetz, Stefan Jachmich, E. Joffrin, Mengdi Kong, M. Lehnen, C. Maggi, Earl S Marmar, E. Nardon, P. Puglia, U. Sheikh, Daisuke Shiraki, S. Silburn","doi":"10.1088/1741-4326/ad1d10","DOIUrl":"https://doi.org/10.1088/1741-4326/ad1d10","url":null,"abstract":"\u0000 Precise values for radiated energy in tokamak disruption experiments are needed to validate disruption mitigation techniques for burning plasma tokamaks like ITER and SPARC. Control room analysis of radiated power (Prad) on JET assumes axisymmetry, since fitting 3D radiation structures with limited bolometry coverage is an underdetermined problem. In mitigated disruptions, radiation is toroidally asymmetric and 3D, due to fast-growing 3D MHD modes and localized impurity sources. To address this problem, Emis3D adopts a physics motivated forward modeling (\"guess and check\") approach, comparing experimental bolometry data to synthetic data from user-defined radiation structures. Synthetic structures are observed with the Cherab modeling framework and a best fit chosen using a reduced χ2 statistic. 2D tomographic inversion models are tested, as well as helical flux tubes and 3D MHD simulated structures from JOREK. Two nominally identical pure neon shattered pellet injection (SPI) mitigated discharges in JET are analyzed. 2D tomographic inversions with added toroidal freedom are the best fits in the thermal quench (TQ) and current quench (CQ). In the pre-TQ, 2D reconstructions are statistically the best fits, but are likely over-optimized and do not capture the 3D radiation structure seen in fast camera images. The next-best pre-TQ fits are helical structures that extend towards the high-field side, consistent with an impurity flow under the magnetic nozzle effect also observed in JOREK simulations. Whole-disruption radiated fractions of 0.98 +0.03/-0.29 and 1.01 +0.02/-0.17 are found, suggesting that the stored energy may have been fully mitigated by each SPI, although mitigation efficiencies well below ITER and SPARC requirements for high energy pulses are still within the large uncertainties. Emis3D is also used to validate JOREK SPI simulations, and confirms improvements in matching experiment from changes to impurity modeling. Time-dependent toroidal peaking factors are calculated and discussed.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"2 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139439441","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-09DOI: 10.1088/1741-4326/ad1c93
Adriana Gabrielle Ghiozzi, Mervi Mantsinen, Pol Pastells, D. Spong, A. Melnikov, Leonid E Eliseev, S. Sharapov
Alfvénic activity has been observed in the TJ-II stellarator which resembles the frequency sweeping demonstrated by Alfvén cascade modes in tokamaks. A numerical validation study was conducted using a reduced magnetohydrodynamic (MHD) model to show that such modes could only have been observed in discharges where the rotational transform profile was non-monotonic. During experiments, coil current was varied which resulted in shifting of the minimum value of the rotational transform profile. To mimic this effect, we study the Alfvénic activity predicted by the reduced MHD model for a set of input rotational transform profiles with varying minima. A mode is found whose toroidal and poloidal mode numbers match those predicted in experiments which sweeps downward/upward in frequency as the minimum value of the rotational transform profile is increased/decreased. The results serve as a demonstration of the validity and utility of magnetohydrodynamic spectroscopy.
{"title":"Modeling of frequency-sweeping Alfvén modes in the TJ-II stellarator","authors":"Adriana Gabrielle Ghiozzi, Mervi Mantsinen, Pol Pastells, D. Spong, A. Melnikov, Leonid E Eliseev, S. Sharapov","doi":"10.1088/1741-4326/ad1c93","DOIUrl":"https://doi.org/10.1088/1741-4326/ad1c93","url":null,"abstract":"\u0000 Alfvénic activity has been observed in the TJ-II stellarator which resembles the frequency sweeping demonstrated by Alfvén cascade modes in tokamaks. A numerical validation study was conducted using a reduced magnetohydrodynamic (MHD) model to show that such modes could only have been observed in discharges where the rotational transform profile was non-monotonic. During experiments, coil current was varied which resulted in shifting of the minimum value of the rotational transform profile. To mimic this effect, we study the Alfvénic activity predicted by the reduced MHD model for a set of input rotational transform profiles with varying minima. A mode is found whose toroidal and poloidal mode numbers match those predicted in experiments which sweeps downward/upward in frequency as the minimum value of the rotational transform profile is increased/decreased. The results serve as a demonstration of the validity and utility of magnetohydrodynamic spectroscopy.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"58 49","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139441633","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：