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.     

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.     

Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead.     

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.     

Calculation of the Heat Flux on the First Wall During Disruption on Tokamak

Disruptions are the most dangerous instabilities in tokamak plasma. During plasma disruption, the large amounts of energy will be deposited on Plasma Facing Components (PFCs) which is a damaging threat for the divertor target and the first wall materials. Therefore, studying the characteristic of heat deposition on the first wall is very significant. The Infrared (IR) camera is an effective tool to measure the surface temperature profile on the first wall on the Experimental Advanced Superconducting Tokamak (EAST). With a finite difference method, the heat flux arrived to the divertor can be calculated from the surface temperature. However, the surface layer on the divertor has a great influence on the calculation of the heat flux on the divertor. The numerical method for solving heat conduction for semi-infinite model is given in this paper. And the thermal resistance of surface layers is considered in this numerical method. In addition, the distribution of heat flux on the divertor during disruption is also shown.     

Evaluation of heat transfer by sublimation for the application to the divertor heat sink for high fusion energy conversion

Thermal and structural responses of divertor target were evaluated by using finite element method. High heat flux simulating ELMs at the level of 100 MW/m² was assumed onto the tungsten armor, and surface temperature profile was obtained. When dynamic heat load over 100 MW/m² was applied, the maximum surface temperature exceeded 1300 °C, and it caused recrystallization of tungsten regardless of the heat transfer below it. The result was used to conduct dynamic heat load experiment on tungsten, and material behavior of tungsten was evaluated under dynamic heat load. This study also proposed new concept of divertor heat sink which can distribute high heat flux and transfers the heat to high temperature medium. It consists of tungsten armor, composite enhanced with high thermal conductivity fiber, and heat transport system applying phase transition. High heat flux simulating ELMs was also applied to target surface of the divertor, temperature gradient, thermal stress of tungsten and composite were evaluated. Based on the results of analysis, thermal structural requirement was considered.     

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.     

Improvement of divertor triple probe system and its measurements under full graphite wall on EAST

Full graphite wall of experimental advanced superconducting tokamak (EAST) has been developed in the spring of 2008. A new divertor triple probe diagnostics system (DTPDs) is built for EAST during this upgrade. The tip shape and connected structure of the probe are optimized for variational magnetic field directions and DTPDs maintenance. The experiment has been carried out with a full graphite wall for EAST, and near double-null diverted plasma is achieved successfully. The evolutions of electron temperature, density, particle flux and power densities along the divertor targets have been obtained with DTPDs.     

Calculation of the heat flux in the lower divertor target plate using an infrared camera diagnostic system on the experimental advanced superconducting tokamak

During the discharging of Tokamak devices,interactions between the core plasma and plasma-facing components(PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared(IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak(EAST) to monitor the surface temperature of the lower divertor target plate(LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX; this is important for future work on related physical processes and heat flux control.     

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation, detachment and redistribution of heat flux, etc. Two sets of probe arrays including 274 probe tips were placed at two ports (approximately 180° separated toroidally), and the spatial and temporal resolutions of this measurement system could reach 6 mm and 1 μs, respectively. A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time. Meanwhile, the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station. The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density, electron temperature, particle flux as well as heat flux during the ELMy H-mode discharges.     

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.     

Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs

The experimental evaluation of the divertor plasma facing components (PFCs) lifetime under transient events, such as edge localized modes (ELMs) and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events is here presented. The experiments have been performed in the frame of an EU/RF collaboration. For carbon fiber composite material the erosion is caused by PAN fiber damage whilst the erosion of tungsten is determined by the melt layer movement and crack formation. The conclusion of this study is that, in addition to the structural change produced in the armor materials by ELMs-like loads, some mock ups showed also a degradation of the thermal fatigue performances     

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.     

Observations of hydrocarbon film deposition in the MAST tokamak

Two divertor tiles from the Mega-Ampere Spherical Tokamak (MAST) were analysed using interference fringe analysis, nuclear reaction analysis and Rutherford backscattering analysis. The analysis allowed the quantification of the co-deposited layers thickness of the deuterium and carbon mixture and the detection of the presence of other impurities such as oxygen and iron. A layer thickness varying from 10 to 270 nm was measured with a typical D/(D+C) ratio of ≈0.36. During the same campaign in which the analysed divertor tiles were exposed to the plasma, the ion flux to the divertor plates has been measured by a set of Langmuir probes mounted on the divertor ribs. This paper shows that the sputtering co-efficient that can be inferred from the ion fluences measured by the probes and by the redeposition of C in those amorphous layers is consistent with physical and chemical sputtering yields in conventional tokamaks. This paper reports the beginning and the intended developments of a project undertaken in MAST to study and quantify the phenomena of plasma surface interaction both at the walls and at the divertor of the machine in order to contribute to the material choice and design of the plasma facing components of a next step device such as ITER.     

Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m² with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m² were performed.Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected during ELMs in ITER gives a first impression of the possible severe material degradation at the surface during operational scenarios at the divertor strike point.     

Overview of HIT-SI Diagnostic Systems

The Helicity Injected Torus with Steady Inductive current drive (HIT-SI) device is a spheromak which uses two inductively driven helicity injectors to provide constant helicity injection. The HIT-SI diagnostic set includes 96 3-D magnetic surface probes, 20 toroidal flux loops, an internal array of three stems of 3-D magnetic probes, a far infrared interferometer, a SPRED diagnostic, two tangentially viewing vacuum ultraviolet (VUV) spectrometers, a bolometer, a Zeff diagnostic, H-alpha detectors, a time integrated CCD camera, and Ion Doppler Spectroscopy (IDS). In addition, a multi point Thomson Scattering system, a microwave interferometer/polarimeter, injector voltage probes, an edge Langmuir probe array, and soft X-ray camera are under development.     

Research plan for divertor simulation making use of a large tandem mirror device

We are planning to start a study of divertor simulation under the closely resemble to actual fusion plasma environment making use of the advantage of open magnetic field configuration and to contribute the solution for realizing the divertor in ITER as a future research plan of Plasma Research Center of the University of Tsukuba. In the research plan, the concepts of two divertor devices are introduced. One has an axi-symmetric divertor configuration with the separatrix which is similar to toroidal divertor of torus systems and the other is a high heat flux divertor simulator by using an end-mirror exit of the existing tandem mirror device. Development of magnetic field configuration for ensuring the MHD stability is under way and a designed example is investigated under the optimal condition for plasma production. Consideration of plasma heating scheme using Fokker-Planck simulation code was successfully performed at both axi-symmetric divertor and end-mirror regions. Preparative experiments using calorimeter, Mach probe and high-speed camera have been started at the end-mirror region and the heat flux density of the level in 1-10 MW m⁻² was achieved in standard hot-ion mode plasma-confining experiments, which gives a clear prospect of generating the required heat flux density for divertor studies.     

Numerical Simulation and Experimental Identification of Divertor Configuration in the HL-2A Tokamak

Single- and double-null divertor configurations in HL-2A are simulated by SWEQU equilibrium code. Lower divertor discharges in the first physics campaign have been achieved by two kinds of power supply method of multipole-field coils. Single-null divertor configuration has been identified by visible photography, target probe arrays and the reconstructed magnetic surface. Magnetic separatrix and minor radius of plasma column are obtained by a reconstructed code of multiple current filaments using 18 Mirnov signals.     

On the design of liquid metal divertors

The divertor technology has become the focus of concern for prospective steady state tokamak reactors. The imposed heat flux and particle flux conditions cast doubt on the feasibility of any solid surface divertor. The aim of this work is to evaluate the existing concepts of liquid metal divertors from the physics, engineering, and safety points of view. Lithium metal is found to relatively suffer from a large tritium inventory that may develop in the form of solid hydride LiH, in addition to the fire hazard potential. Gallium does not form hydride within the temperature range of interest and is inactive with oxygen, therefore it is considered a more favorable metal. The flowing film and pool type divertors are found to suffer from the possible blistering erosion and resulting plasma contamination. The self-cooled liquid metal film divertor suffers also from a linear MHD instability, in addition to complicating factors such as the variation of the liquid metal electric property, dynamics of halo current, and the possible incompatibility with the insulator coating. The liquid gallium droplet curtain divertor appears to be the most feasible and roust, but its high temperature vapor is corrosive to most structural materials.     

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.