, Volume 1, Issue 2, pp 122–131 | Cite as

Plasma–tungsten interactions in experimental advanced superconducting tokamak (EAST)

  • Fang Ding
  • Guang-Nan LuoEmail author
  • Xiahua Chen
  • Hai Xie
  • Rui Ding
  • Chaofeng Sang
  • Hongmin Mao
  • Zhenhua Hu
  • Jing Wu
  • Zhen Sun
  • Liang Wang
  • Youwen Sun
  • Jiansheng Hu
  • the EAST Team
Review Paper


Tungsten (W) is used as the armor material of the International Thermonuclear Experimental Reactor (ITER) divertor and is regarded as the potential first wall material of future fusion reactors. One of the key challenges for the successful application of W in fusion devices is effective control of W at an extremely low concentration in plasma. Understanding and control of W erosion are not only a prerequisite for W impurity control, but also vital concerns to plasma-facing component (PFC) lifetime. Since the application of ITER-like water-cooled full W divertor in EAST in 2014, great efforts were made to investigate W erosion by experiment and simulation. A spectroscopic system was developed to provide a real-time measurement of W sputtering source. Both experiment and simulation results indicate that carbon (C) is the dominant impurity causing W sputtering in L-mode plasmas, which comes from the erosion of C plasma-facing material (PFM) in the lower divertor and the main chamber limiters. The mixture layer on the surface of W PFCs formed through redeposition or the wall coating can effectively suppress W erosion. Increasing the plasma density and radiation can reduce incident ion energy, thus alleviating W sputtering. In H-mode plasmas, control of edge localized mode (ELM) via resonant magnetic perturbation (RMP) proves to be capable of suppressing intra-ELM W erosion. The experiences and lessons from the EAST W divertor are beneficial to the design, manufacturing and operation of ITER and beyond.


Divertor Tungsten sputtering Erosion EAST Spectroscopy 

1 Introduction

The divertor is one of the key components in future fusion reactors, which plays a major role in removing the huge heat and particle fluxes from fusion plasmas, screening the impurity influx generated from plasma facing surfaces, and exhausting the helium (He) ash produced by the deuterium (D)–tritium (T) reactions. There are great challenges for the plasma-facing material and component (PFMC) used as the divertor in current tokamaks and future reactors. For example, for International Thermonuclear Experimental Reactor (ITER) being built in Cadarache, France, the divertor is to experience steady-state particle fluxes of ~ 1024 m−2 s−1, the heat load of ~ 10 MW m−2, and the transient heat load up to 1–10 GW m−2 [1, 2]. However, there are no materials fully satisfactory now. Due to its favorable properties such as high melting temperature, high thermal conductivity, low sputtering yield and low retention with hydrogen isotopes, pure tungsten (W) is chosen by ITER as PFM for the whole divertor (dome and both vertical targets) [3, 4]. Plasma wall interaction (PWI) is a great concern in achieving the high-performance and long-pulse discharges as well as the lifetime of the PFMC. A strong W atom flux could be produced on the divertor surface via erosion processes by the impinging impurities and energetic fuel ions (H, D, T). The W erosion would affect the lifetime of PFMC. Moreover, the eroded W particles can be transported into the core plasma and cause strong radiation, which inevitably cools down the plasma and degrades the plasma performance. If the critical W concentration of 10−5 in the core plasma is exceeded, the fusion process may not be achieved in a reactor [5]. Therefore, good understanding and control of the W erosion are the most important for the successful application of W in fusion devices.

2 Application of tungsten as plasma-facing materials in EAST

The experimental advanced superconducting tokamak (EAST) is a Chinese tokamak operated by Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) since 2006, with both toroidal and poloidal superconducting coils and flexible divertor configurations (single and double nulls). EAST is capable of simulating ITER operation scenarios, thus contributing significantly to ITER engineering issues and physics understanding. ASIPP developed a batch manufacturing technology for ITER-like water-cooled full W-PFC which employs copper alloy (CuCrZr) as a heat sink material and the hot isostatic pressing (HIP) method to join the W-PFM and the CuCrZr heat sink together [6]. Therewith ASIPP upgraded the EAST upper divertor from the bolted graphite tile PFC [7] into the ITER-like water-cooled full W-PFC in 2014, and the W divertor came into service successfully [8, 9, 10, 11]. Fully non-inductive steady-state H-mode plasmas (H98, y2 ~ 1.1) were achieved with a duration of more than 60 s in 2016 experimental campaign [10] and further extended to ~ 100 s in 2017 campaign with a good control of impurity and heat exhaust benefiting from the upper W divertor [11]. A new lower W divertor is designed [12] and will replace the existing graphite divertor in 2020. The new lower divertor has a close outer divertor and a comparably open inner divertor to achieve more balanced detachments at outer and inner targets to optimize the power handling capability and to improve compatibility with good plasma performances in long-pulse discharges. The engineering structure of the new lower divertor is also optimized to enhance the reliability and accommodate flexible plasma configurations. During these campaigns, the plasma–tungsten interactions were monitored and analyzed, which greatly promote our understanding of the underlying mechanisms, especially for W impurity production and its control.

3 W sputtering in EAST

3.1 In-situ monitoring of W sputtering source

The W influx comes from sputtering of surfaces where the plasma contacts, i.e. divertor targets in EAST. The sputtered W atoms are excited by plasma in the immediate vicinity of the targets. The line radiations from the excited plasma can be measured spectroscopically along a line-of sight (LOS) directed at the sputtering surface. These neutral W atoms tend to be well localized around the target surfaces before they are ionized. The measured radiances from excited neutral W atoms can, therefore, be converted into W atom fluxes from the targets.

A multichannel spectroscopic system was built on EAST to dedicatedly measure the emission lines of impurities in the upper W divertor region [13]. This system has a vertical array of 22 LOSs directed at the outer target and 17 LOSs at the inner target, respectively, which provides profile measurements of W sources along both divertor targets with a 13 mm poloidal resolution (Fig. 1). Two types of spectroscopic detecting systems are used for the analysis of the plasma emission light transmitted through fibers. The spectrometer and electron-multiplying charge-coupled device (EMCCD) provide the multichannel spectral analysis with high spectrum resolution. The photomultiplier tube and narrow bandpass filter enable fast tracking of the effects of transient events, such as edge localized mode (ELMs). The so-called S/XB method [14, 15] is applied to quantify the impurity flux in the W divertor including the sputtered W atom flux and C impurity flux.
Fig. 1

a Poloidal lines of sight of the multichannel spectroscopy system in the upper W divertor on EAST; b photos of the upper W divertor with circles on the surface indicating the locations of LOSs (left).

Reproduced with permission from Ref. [13]. Copyright 2017 AIP Publishing

Figure 2 shows a poloidal profile of sputtered W atom fluxes on the outer divertor target in an L-mode discharge sustained by modulated neutral beam (NBI) and lower hybrid wave (LHW) heating. The plasma current is 0.4 MA and the line-averaged plasma density is 2 × 1019 m−3. The distribution of the probe ion saturation current density on the target [16] was also given for comparison with the W atom flux. The vertical coordinate indicates the poloidal distance from the upper edge of the outer target schematically shown on the right. The W atom influx exhibits a clear association with the ion flux onto the divertor target, both spatially and temporally. The highest W atom flux appears around the strike point similar to the probe ion flux. In the incident ion flux onto targets, impurities play an important role in the production of W influx, since the energy threshold of W sputtering by D ions is high (~ 250 eV) [17, 18, 19] and cannot be reached at the divertor targets in normal L-mode discharges. In Fig. 3, the W atom flux increased during the NBI heating phase, well corresponding to the variation of low atomic number (Z) impurities C and Si, etc., despite the relatively stable electron temperature Te (Fig. 3d). The increased radiation power (Fig. 3b) could be partly due to elevated impurity level during the NBI phase.
Fig. 2

Poloidal distribution and temporal evolution of W atom fluxes (left) and the probe ion saturation current density (middle) on upper outer divertor target in shot 59390. On the right is a schematic of the outer target approximately corresponding to the vertical coordinate of the two diagrams on the left.

Reproduced with permission from Ref. [13]. Copyright 2017 AIP Publishing

Fig. 3

Time traces of a power injection, b plasma stored energy and total radiation power, c main impurity fluxes in the upper outer divertor obtained through measuring the spectral emission lines and d probe ion saturation current density and electron temperature around the striking point in an L-mode divertor discharge of shot 59390.

Reproduced with permission from Ref. [13]. Copyright 2017 AIP Publishing

3.2 Modeling of W sputtering mechanism

A large portion of EAST wall is covered with carbon (C) tiles, including the graphite lower divertor, neutral beam shinethrough region at the high field first wall and the guarder limiter for radio frequency heating antennas. The C impurity in plasma comes from the erosion of these C tiles and contributes greatly to the W erosion at the upper W divertor targets. The simulations by using the SOLPS5.0/EIRENE99 code package were carried out by Sang et al. [20] to study C impurity transport and the effects on the W erosion at the W divertor target in EAST. The results show that C impurities produced mainly through chemical sputtering [21] at the bottom C targets could be transported into the upper divertor in both upper single-null (USN) and double-null (DN) configurations. But the dominant C ionization states responsible for the W erosion differ in USN and DN configurations.

In the USN configuration, the highest eroded W fluxes at upper outer divertor target are caused by C4+ and C6+ incidence (Fig. 4a). In the DN configuration, the W target is mainly eroded by C6+ as shown in Fig. 4b. The primary reason could be related to the different proportions of C4+ and C6+ ion fluxes in the incident C ions (Fig. 4c, d). In the simulation for the USN configuration, only neutrals can reach the lower graphite divertor. For eroded C particles in the lower divertor, the cross-field transport into core plasma is not strong, while the transport path through the scrape-off layer (SOL) to the upper divertor becomes more important, where Te is ~ 100 eV, which is too low to ionize C ions to C6+. In the DN configuration, the plasma can directly contact the lower C divertor targets and the eroded C particles can be rapidly ionized. After penetrating into the core, most of these C ions can be further ionized to the highest charge state C6+. Some C6+ ions are finally transported out of the separatrix and deposited either on the lower divertor targets or on the upper divertor targets. So the C6+ ion flux is much higher than other C ion fluxes onto targets as shown in Fig. 4a. Moreover, the ions with a higher charge state can obtain a higher incident energy through the plasma sheath acceleration. This is another important reason for C6+ ion fluxes to dominate the W sputtering at the upper target.
Fig. 4

Calculated W erosion rate by C ions at the upper outer target a in the USN configuration and b in the DN configuration; calculated C ion fluxes at the upper outer target c in the USN configuration and d in the DN configuration. The total particle flux profiles along the upper outer target are presented in the inset-graphs.

Reproduced with permission from Ref. [20]. Copyright 2018 AIP Publishing

The W sputtering rate depends on both the incident ion energy and flux. The W erosion rate at divertor targets can be reduced by the increasing plasma density at the edge of main plasma. In Fig. 5, a typical simulation shows that as the upstream density rises from 9.3 × 1018 m−3 to 2.6 × 1019 m−3, the peak electron temperature drops from 88 to 11 eV, and the peak W erosion rate decreases apparently from 3.3 × 1017 to 2.3 × 1016 W atoms m−2 s−1. Moreover, the high collisionality at a high upstream density can slow down the parallel velocities of the impurity ions towards the divertor targets, reducing C impurity fluxes reaching the outer target of upper W divertor [20].
Fig. 5

Dependence of the peak W erosion rate at the UO target on the plasma density at middle plane separatrix. The dependence of corresponding peak electron temperature on the plasma density is also presented in the inset graph.

Reproduced with permission from Ref. [20]. Copyright 2018 AIP Publishing

The 3D Monte Carlo code ERO was also used to improve the understanding of W sourcing mechanisms by simulating W sputtering and re-deposition on the EAST W divertor target [22]. The photon flux of W atom emission line (400.9 nm) was simulated. Its poloidal distribution at the upper outer (UO) divertor target was compared with the spectroscopic measurements in an attached L-mode plasma in EAST as shown in Fig. 6. A reasonable match between the experimental measurement and simulation result can be obtained only if the returned eroded impurities are taken into account in the simulation with the carbon concentration ρc assumed in the range of 0.6–1%. In the strike point region, the eroded impurities including C and W mostly return and greatly enhance the gross W erosion rate. The large deviation in the regions far away from the strike point could result from the measuring errors of divertor probes and the spectroscopic system as well as the assumption of constant C content in the incident plasma flux in the simulation.
Fig. 6

Poloidal profiles of W I (400.9 nm) photon fluxes at different C percentages simulated by ERO. Corresponding spectroscopic measurement is also presented for comparison.

Reproduced with permission from Ref. [22]. Copyright 2018 AIP Publishing

The eroded neutrals, once ionized, would return to the divertor surface during their first gyration in the magnetic field, which is called “prompt re-deposition”. This mechanism not only determines the net erosion of the PFMs, the difference between gross erosion and re-deposition, but also changes the composition of the divertor surface, which profoundly influences the gross erosion rate. The ERO modelling results indicate that W redeposition ratio is ~ 95% under the plasma conditions in Fig. 6. The modeled C re-deposition ratio is 62%, resulting in a high C concentration in the surface interaction layer even with a small C content in the incident plasma flux. The increasing C content in the plasma flux can first enhance W erosion through increasing the C ion bombarding the divertor surface (Fig. 7). Concurrently, the C concentration in the surface interaction layer also rises, which results in the reduction of the W gross erosion rate. It can be seen in Fig. 7 that a rollover in the W gross erosion rate appears with the increase of the C content in the plasma, indicating the suppression of W erosion if the C content is further increased. The simulations also show that the W gross erosion rate presents a noticeable increase if the contribution of returned eroded impurities to W sputtering is considered. Overall, net C deposition appears at the dome plate and part of the vertical target close to the dome plate, whereas net W erosion happens in the region around the strike point at the target plane.
Fig. 7

Dependence of the C percentage in the surface interaction layer and W gross erosion rate on the C content in the background plasma flux in equilibrium simulated by ERO.

Reproduced with permission from Ref. [22]. Copyright 2018 AIP Publishing

3.3 Control of W sputtering source

A good control of the W source is vital for maintaining an accepted W concentration in the core plasma, which is a prerequisite for the achievement of long pulse H-mode plasmas with good confinement. Gas puff can effectively suppress the W sputtering by reducing the electron temperature in the edge plasma. Figure 8 presents an L-mode USN discharge in which 3 D2 supersonic molecule beam injection (SMBI) pulses were injected at gradually increased puff rates, respectively, at 4.0 s, 4.9 s and 5.9 s indicated by three yellow lines. This discharge was sustained by 2.45 GHz LHW (0.4 MW) and 4 GHz LHW (2 MW) with a plasma current of 0.4 MA and toroidal magnetic field of 2.3 T. It can be seen that each puff causes a rise of the line-averaged electron density at mid-plane and the sputtered W atom flux at the divertor target also presents a drop correspondingly. Figure 8 c plots the sputtered W atom flux against Te at the UO divertor target measured by the embedded probes. The W atom flux drops distinctly with the decrease of Te due to D2 gas puff. At the energy of ~ 15 eV, the W atom flux drops to a low noise level. The W sputtering threshold energy for deuterons is ~ 250 eV due to their low masses [17, 18, 19]. The deuterons with impact energies lower than 250 eV do not contribute to W impurity production by physical sputtering. In L-mode discharges, the physical sputtering rate of tungsten is generally governed by impurity composition, their energies and fluxes if their energies exceed the threshold. Moreover, the mixture deposition layer on the W surface due to wall coating and redeposition should be an important element influencing the W erosion rate.
Fig. 8

Time evolutions of a the line averaged electron density and b sputtered W atom flux at the outer divertor target; c sputtered W atom flux in b plotted against Te at the target. D2 gas was injected at the outside mid-plane through SMBI at 4 s, 4.9 s and 5.9 s, respectively, with gradually increased injection rates

Both Li and Si wall conditionings are applied in EAST to coat the first wall and divertors. The Si coating is implemented at night without normal plasma discharges. The vacuum chamber is filled with the mixture of silane (SiD4) and He at a ratio of 1:9. SiD4 molecules are dissociated in the ion cyclotron resonance frequency (ICRF) discharge and then the released Si atoms deposit on the wall surface. The Li wall coating is obtained by heating Li powder in the vacuum chamber to form Li vapor almost every night. Li particles are dispersed through glow discharge and finally coated on the wall. The effects of two kinds of coatings on the suppression of W sputtering were compared. Figure 9 presents the dependences of typical effective W sputtering yields with Li- and Si-coated walls in EAST on the electron temperature at the UO divertor target. The effective W sputtering yield is calculated by normalizing the sputtered W atom flux to the incident probe ion current density. The similar dependences in the Joint European Torus (JET) and Axially Symmetric Divertor Experiment (ASDEX) Upgrade [14, 23] are also included in Fig. 9 for comparison. The solid lines represent the calculated results using the Transport of Ions in Matter (TRIM) code with different impurity percentages in the incident plasma flux. It can be seen that the effective W sputtering yield with the Si wall coating is higher than that with the Li coating, respectively, corresponding to the solid lines of 2% C4+ and approximately 1% C4+. Here the C4+ ion is a representative of the combined effects of impurity (C, O, N, Si, etc.) ions in the incident plasma flux. This reveals that the effect of the Li coating on W sputtering suppression is better than the Si coating. Moreover, it is noticed that EAST has effective W sputtering yields similar to those in ASDEX Upgrade before boronization [23], but higher than those in JET [14]. This could be related to the different wall material configurations. In JET, the lower W sputtering yield can be ascribed to the low impurity level in plasma with the ITER-like wall material configuration, i.e. the beryllium (Be) first wall and W divertor [24]. Be plays a role like Li in EAST and the W sputtering threshold for beryllium is higher than that for C. In EAST, a portion of wall including the lower divertor, antenna guard limiters, etc. is still covered with graphite tiles. C ion bombardment is considered as the main mechanism for W erosion in an L-mode plasma. Despite that, the C impurity concentration can be well controlled at a low level by actively coating walls with Li. Thus, the W sputtering yield in EAST can be maintained at a level comparable to that in ASDEX Upgrade with full W walls where the responsible impurity is the residual C.
Fig. 9

Dependence of effective W sputtering yields on electron temperature at divertor targets in EAST, JET-ITER-like Wall (ILW) [14] and ASDEX Upgrade [23]. Solid lines denote the calculated results using TRIM code with different impurity percentages in the incident plasma flux.

Reproduced with permission from Ref. [25]. Copyright 2016 Elsevier

Besides the intrinsic C impurity, residual oxygen is found to be an important impurity responsible for W sputtering. Divertor impurity levels in the W divertor during 1-day discharges were compared before and just after once exposure to air due to inner component repairing. The impurity levels are indicated by normalizing the impurity emission line intensities to the Dδ line intensity in the upper outer divertor as shown in Fig. 10. It can be seen that the W sputtering yield, indicated by the W I intensity normalized by the Dδ line intensity, significantly increases after the exposure to air despite the high lithium level. This increase can be related to the elevated O II and C II intensities as shown in Fig. 10. In 1-day discharges, C and O signals gradually increased due to the degradation of the Li coating, so did the W I signal. The O impurity not only physically sputters W, but also chemically erodes C walls, and thus increases the C concentration in plasma. Therefore, it is necessary to reduce the O impurity level to the minimum by sufficient wall baking and conditioning.
Fig. 10

1-day evolutions of W, Li, O, C impurity contents in the UO divertor with the Li wall coating before and after exposure to air.

Reproduced with permission from Ref. [25]. Copyright 2016 Elsevier

As another Li wall conditioning method, the real-time Li aerosol injection technique is also developed on EAST and actively supports the achievement of long-pulse ELM-free H modes [26, 27]. The edge impurity levels and particle recycling can be reduced by real-time Li injection, thus improving plasma performances. It is observed that the real-time Li aerosol injection can more effectively suppress the W sputtering at the divertor target compared with the Li evaporation wall coating. Figure 11 presents Li injection effects on divertor conditions in a typical USN divertor discharge. 3 MW LHW was used to sustain the plasma and the central line averaged electron density was ~ 2.9 × 1019 m−3. The Li aerosol dropper system above the upper W divertor was activated at 4.7 s and the Li aerosol was injected into upper divertor plasma at a flow rate of 60 mg s−1. The Li I line intensity started to rise after an ~ 1 s delay (Fig. 11d). Correspondingly the perpendicular heat flux, sputtered W influx, spectroscopic C II line intensity and Te at the upper outer divertor target dropped apparently (Fig. 11). These should reflect that the injected Li aerosol can enhance the divertor radiation, dissipate the particle energy and cool the plasma in the divertor, which is confirmed by the absolute extreme ultraviolet (AXUV) signal as shown in Fig. 11e. Furthermore, it is observed that the W source at the divertor target is kept at a low level in the subsequent discharges for several days even without further Li aerosol injection. This reveals that the real-time Li aerosol injection can more efficiently coat the plasma-contacted region, e.g. the divertor target planes, as the injected Li is entrained in the motion of the plasma flux.
Fig. 11

Li aerosol injection effects on the UO divertor plasma in an L-mode discharge. a Perpendicular heat flux to the UO divertor target obtained by the triple probes embedded in the target; b eroded W atom flux at the UO divertor target; c electron temperature at UO divertor target; d photon fluxes of W I, C II and Li I line emissions, in which the line intensity of W I is multiplied by 10; e uncalibrated plasma radiation in upper divertor measured by the fast AXUV system.

Reproduced with permission from Ref. [25]. Copyright 2016 Elsevier

In H-mode plasmas, ELM bursts can cause much stronger W sputtering at the divertor target than that in L-mode plasmas due to the high power and particle flux from the edge plasma during ELMs. Spectroscopic measurements show that W erosion during ELMs accounts for ~ 70% of the total W erosion in H mode discharges. Intra-ELM W sputtering shows a different dependence on the plasma parameters, since the intra-ELM sputtering may be linked to pedestal parameters rather than edge conditions [28]. In recent years, 3D Magnetic Perturbation (MP) fields are investigated in many tokamaks. MP fields break the toroidal axisymmetry of the magnetic configuration and lead in effect to a change of the radial plasma transport, which is considered as an important means to suppress or mitigate the ELMs and will be applied in ITER. In EAST, intra-ELM W sputtering can be suppressed by applying resonant magnetic perturbation (RMP) as shown in Fig. 12. With the rise of the RMP current, the probe ion saturation current at the divertor target presents a descending trend, while the W I line intensity during ELMs keeps relatively stable until ELMs are completely suppressed after the RMP current reaches the top value of 2.2 kA turn (kAt). Meanwhile, 3D characteristic of impurity sources is recently observed in EAST, which could be due to the 3D power and particle flux footprints on divertor targets [29]. The 3D field effects on PWI in the divertor are not very clear up to now and the relevant investigation is still ongoing.
Fig. 12

Suppression effects of n = 2 RMP on intra-ELM W sputtering (top) and intra-ELM incident saturation ion flux (bottom) at the UO divertor target

4 Summary and outlook

The poloidal distribution and temporal evolution of tungsten sputtering sources at the W divertor targets of EAST are quantified via real-time spectroscopic measurements of the emission line of sputtered W atoms. The strongest W source appears around the strike point due to the peak incident ion flux and particle energy. C is considered as the dominant impurity to cause W sputtering in L-mode plasmas, which could be produced by erosion of graphite tiles in the lower divertor and other C-PFCs in EAST. Simulation results show that C6+ and C4+ are the main ions contributing to the W sputtering flux due to their high fraction in the incident ion flux and high incident energy through the plasma sheath acceleration. The two kinds of C ions also vary in their relative proportions in USN and DN configurations, which could be explained by different transport paths of C impurities from the lower C divertor to the upper W divertor. Increasing the C concentration in the plasma flux can first enhance and then suppress W erosion through a C/W interaction layer on the surface. The contribution from returned erosion particles to W erosion cannot be neglected in explaining the experimental observation. W sputtering can be suppressed through increasing the plasma density and radiation, thus reducing particle bombarding energy and lowering low Z impurity concentration by Li wall conditioning. For ELM-induced W sputtering, effective ELM control will be mandatory.

ITER will start with a full W divertor from its initial phase on, but until now world fusion society lacks of mature technology for batch manufacturing of W-PFCs. Furthermore, long-pulse and high-performance plasmas with a full W divertor configuration are rarely investigated due to lack of qualified tokamaks. ASIPP realized batch manufacturing of the W-PFCs for EAST, achieved the full W upper divertor in 2014 and planned to upgrade the lower divertor into full W PFCs in 2020. The experience and lessons learnt from these engineering activities, together with the physics understanding under full W divertors will be much beneficial to the design, manufacturing and operation of ITER.



This work was supported by National Natural Science Foundation of China (NSFC) (Grant No. 11575243), the National Key Research and Development Program of China (Grant Nos. 2017YFE0301300, 2017YFA0402500), and the Users with Excellence Project of Hefei Science Center CAS (Grant No. 2018HSC-UE008).


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Copyright information

© The Nonferrous Metals Society of China 2019

Authors and Affiliations

  • Fang Ding
    • 1
  • Guang-Nan Luo
    • 1
    Email author
  • Xiahua Chen
    • 1
  • Hai Xie
    • 1
  • Rui Ding
    • 1
  • Chaofeng Sang
    • 2
  • Hongmin Mao
    • 1
  • Zhenhua Hu
    • 1
  • Jing Wu
    • 1
  • Zhen Sun
    • 1
  • Liang Wang
    • 1
  • Youwen Sun
    • 1
  • Jiansheng Hu
    • 1
  • the EAST Team
  1. 1.Institute of Plasma PhysicsChinese Academy of SciencesHefeiChina
  2. 2.Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of PhysicsDalian University of TechnologyDalianChina

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