Study of Striated Heat Flux on EAST Divertor Plates Induced by LHW Using Infrared Camera

An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma facing components. Modified magnetic topology induced by lower hybrid wave （LHW） can lead to the formation of striated heat flux （SHF} on divertor plates which can be clearly observed by IR camera. In this paper, LHW power modulation is applied to analyze the appearance of SHF. It is also demonstrated that deuterium （D） pellet injection and supersonic molecular beam injection （SMBI） can to some extent reduce the heat flux on the outer strike point （OSP）, but enhance the SHF on lower outer plates （LOP） of divertor. This may provide an optional approach to actively control the distribution of heat flux on diveror plates, which can protect materials from long duration high-heat flux.     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Study of Divertor Heat Patterns Induced by LHCD L-Mode Plasmas Using an Infra-Red Camera System on EAST

Divertor heat patterns induced by Lower Hybrid Current Drive(LHCD) L-mode plasmas are investigated using an infra-red(IR) camera system on an Experimental Advanced Superconducting Tokamak(EAST). A two-dimensional finite element analysis code DFlux is used to compute heat flux along the poloidal divertor target and corresponding quantities. Outside the Origin Strike Zone(OSZ), a Second Peak Heat Flux(SPHF) zone, where the heat flux is even stronger than that at the OSZ, appears on the lower-outer(LO) divertor plates with LHCD and disappears immediately after switching off the LHCD. The main heat-flux shifts from the SPHF zone towards the OSZ when the divertor configuration converts from double null to lower single null, indicating that the growth of the SPHF zone is apparently affected by a plasma magnetic configuration. The heat patterns on the LO divertor plates are observed to be different from that on the lower-inner(LI) targets as the SPHF zone appears only on the LO divertor target. It is also found that the heat flux at the SPHF zone was obviously enhanced after the Supersonic Molecule Beam Injection(SMBI) pulse.     

Electron density calculation based on Stark broadening of D Balmer line from detached plasma in EAST tungsten divertor

The electron density within the volume of the tungsten divertor of the Experimental Advanced Superconducting Tokamak (EAST) is calculated based on D_ε line (396.9 nm) Stark broadening (SB) measurements. The quasistatic approximation is employed in the SB calculation of the D_ε line. The influences of other broadening mechanisms on the calculation error of electron density have been evaluated. The SB method is applied to the study of spatial distribution and time evolution of the electron density in the W divertor. Two electron density bands are observed in the detached divertor plasma during an L-mode discharge sustained by low hybrid wave (LHW) heating, which could be related to the striated particle flux distribution induced by LHW. After the onset of detachment, the upper electron density band corresponding to outer strike point firstly increases then decreases, while the lower density band corresponding to striated particle flux increases continually although the electron densities from Langmuir Probes at the divertor plate keep a descending trend. This could indicate a downward movement of the radiation region that approximately moves along the magnetic field lines after the onset of detachment.     

Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

Study on divertor heat flux under n = 3 and n = 4 resonant magnetic perturbations using infrared thermography diagnostic in EAST

Resonant magnetic perturbations (RMPs) with high toroidal mode number n are considered for controlling edge-localized modes (ELMs) and divertor heat flux in future ITER H-mode operations. In this paper, characteristics of divertor heat flux under high-n RMPs (n = 3 and 4) in H-mode plasma are investigated using newly upgraded infrared thermography diagnostic in EAST. Additional splitting strike point (SSP) accompanying with ELM suppression is observed under both RMPs with n = 3 and n = 4, the SSP in heat flux profile agrees qualitatively with the modeled magnetic footprint. Although RMPs suppress ELMs, they increase the stationary heat flux during ELM suppression. The dependence of heat flux on ${q}_{95}$ during ELM suppression is preliminarily investigated, and further splitting in the original strike point is observed at ${q}_{95}=4$ during ELM suppression. In terms of ELM pulses, the presence of RMPs shows little influence on transient heat flux distribution.     

Magnetic configuration flexibility of snowflake divertor for HL-2M

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

HL-2M Divertor Geometry Exploration with SOLPS5.0

One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Pack- age is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the di- vertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.     

Reduction of heat flux on divertor plates by remote radiative cooling in doublet III

Using a single null divertor configuration, heat flux intensity and its profile on the divertor plates as a function of plasma current and density were measured with an infrared camera and thermocouples. The vertical width of the heat flux on the divertor plates 2λ is ≈ 10 cm at the lower separatrix and is ≈ 5.5 cm at the upper separatrix. A diffusion coefficient D which is obtained from the measurement of the diffusion length across the scrape-off field lines is roughly proportional to and its magnitude is on the order of Bohm diffusion. The heat flux on the plates decreases by more than a factor of 5 with increasing electron density in the main plasma and is much smaller than that on the limiters in non-diverted plasmas. Only 3% of ohmic input power goes into the divertor plates at high density of the main plasma, while ≈ 20% goes in at low density. The decrease of heat flux is in good agreement with the increase of radiation loss in the divertor region. The heat flux on the divertor plates can be reduced by remote radiative cooling in high density discharges.     

托卡马克超声分子束注入加料的进展

超声分子束注入作为一种新的托卡马克加料方法由作者在1992年首次提出并于当年在中国环流器一号(HL-1)装置演示成功,随后相继应用于中国环流器新一号(HL-1M)和中国科学院超导托卡马克HT-7装置。超声分子束注入等离子体呈现出电子密度峰化和温度中空分布的特征;等离子体流极向旋转速度提高,边缘扰动被抑制,等离子体能量约束得到改善。加料效率较常规脉冲送气提高一倍,而滞留器壁的粒子大为减少。近期开展的高气压氢超分子束注入实验,在束流中发现团簇流,可注入等离子体中心区域。多脉冲分子束注入形成电子密度的阶跃上升,如同冰弹丸注入效果。近年来该项技术已陆续应用于国外大型托卡马克和仿星器,是核聚变装置稳态运行的一种有效的加料方法。     

Temperature Distribution and Heat Flux on the EAST Divertor Targets in H-Mode

An infrared camera (IR) has been put into operation in the Experimental Advanced Superconducting Tokamak (EAST), which is used to measure the temperature distribution on the surface of lower divertor target plates. With a finite di®erence method, the heat flux onto the divertor target plates is calculated from the surface temperature profile. The high confinement mode (H-mode) with type-III edge localized modes (ELMs) has been obtained with about 1 MW lower-hybrid wave power on the EAST in the autumn experiment in 2010. The analyzed H-mode discharges were lower single null X-point diverted discharges with a density range of < ne > (1~ 4)x 10¹⁹ m^-3. The surface temperature of the inner target plate increases with heating power. The peak temperature on the surface of target plates is lower than 200 oC with about 2.4 MW heating power. Comparison among the heat flux profiles occurring in di fferent phases in the same discharge has been erformed. It indicates that the heatflux profile obviously changes from the ohmic phase to the H-mode phase, and the full width at half maximum (FWHM) of the heat flux pro file is the narrowest during the ELM-free H-phase. On the outer target plate, the peak heat flux exceeds 2 MW/m2 during the ELMy H-mode phase, whereas it is only about 0.8 MW/m2 during the ELM-free phase in the same discharge.     

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium (He) discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor. This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges. During density ramp-up, the particle flux shows a clear rollover, while the electron temperature around the outer strike point is decreasing simultaneously. The divertor detachment also exhibits a significant difference from that observed in comparable deuterium (D) discharges. The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power, and increases with the heating power. Moreover, detachment assisted with neon (Ne) seeding was also performed in L- and H-mode plasmas, pointing to the direction for reducing the density threshold of detachment in He operation. However, excessive Ne seeding causes confinement degradation during the divertor detachment phase. The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.     

Numerical studies of the influence of seeding locations on D-SOL plasmas in EAST

Impurity seeding has been found effective for divertor detachment operations and the seeding location plays a key role in this process. In this work, we use the fluid code SOLPS-ITER to study the influence of seeding locations on divertor and scrape-off layer (D-SOL) plasmas in Experimental Advanced Superconducting Tokamak (EAST) with neon seeding. Simulation results indicate that the neon is a highly effective impurity in mitigating the heat flux and electron temperature peaks on the target of the divertor and achieving the partial detachment on both inner and outer targets. Further, by comparing results of the seeding at the private-flux region (PFR) plate (called 'TP' location) and the outer target (called 'XP' location), we find that the impurity density and power radiation for TP case are higher in core and upstream regions and lower in the divertor region than that for seeding at the XP, and the difference becomes more and more obvious as the seeding rate increases. It clearly demonstrates that the seeding at the XP location is more appropriate than at the TP location, especially in high seeding rate conditions.     

Efforts Made to Handling Steady-State High Heat Flux in EAST

There have been three generations divertor designed for EAST to handle steady-state high heat flux form plasma. The first generation divertor was used on the initial phase of the plasma burning. The first generation divertor was just stainless plate 5 mm in thickness bolted on supports which had been applied since 2006–2007. From 2008 to 2013 the second generation divertor has been used. The second generation divertor was graphite divertor that consisted of graphite tiles, heat sink (CuCrZr) and supports (316L). The third generation divertor was tungsten divertor with ITER like design that had been used science 2014. Now days the upper divertor is tungsten divertor (80 modules) and the lower divertor is graphite divertor (16 modules) in EAST. Tungsten divertor is able to withstand 10 MW/m² heat flux on its strike point and graphite divertor can bear 2 MW/m² under same conditions. It is very important to make every efforts to improve thermal extraction technology of divertor by comparing and practice different designs. Such efforts made in EAST can bring experiences and answers for ITER or any next divertor fusion device on nuclear phase.     

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.     

Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon–fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.