Plasma–tungsten interactions in experimental advanced superconducting tokamak (EAST)
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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.
KeywordsDivertor Tungsten sputtering Erosion EAST Spectroscopy
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 . 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 . Therewith ASIPP upgraded the EAST upper divertor from the bolted graphite tile PFC  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  and further extended to ~ 100 s in 2017 campaign with a good control of impurity and heat exhaust benefiting from the upper W divertor . A new lower W divertor is designed  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.
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.  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  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.
3.3 Control of W sputtering source
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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