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Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s −1 upto 12 g s −1 depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping. Extreme heat loads are predicted on future fusion reactor divertorsExpected engineering lifetime of plasma-facing components (PFC) loaded by 16 MW m −2 corresponds to an acceptable 3 months of cumulated plasma exposure (figure 18 in [1]), however, already for 25 MW m −2 it's only an hour (totally unacceptable). We predicted [2, 3] plasma heat flux q ⊥ perpendicular to the surface of divertor tiles (far off its edges) in ITER attached L-mode (without any fusion) as already q ⊥ =10 MW m −2 . Q=10 ITER H-mode yields P divertor ITER =(1-f rad )100 MW, ⅔ of which deposits on the outer target A dw =2πR 0 f x λ q,integral =0.4 m 2 area [4,42]. Properly controlled impurity seeding increases natural radiation fraction up to f rad =85% [5] on present tokamaks without undesired cooling of the fusing hot plasma core. Accounting further for the toroidal bevel, one gets q ⊥ =⅔P divertor ITER /A dw 4.2°/2.7°=31 MW m −2 [6], thus twice above the engineering limit.Assuming that rather the very complex and recent turbulence models XGC1 [7] and BOUT++ (figure 3 in [Xu19]) will be closer to reality than these empirical scalings, the energy will be deposited on a much more optimistic 5-10× larger area (λ q =6 mm thanks to stronger SOL turbulence due to larger a/r ion Larmor ), predicting thus 5<q ⊥ [MW m −2 ]<16 [1]. The European DEMO 2 GW fusion power plant study [8, 9] counts with similar P SOL =150 MW due to assumed strong core line radiation, namely to stay within the 16 MW m −2 limit, the DEMO edge+SOL+divertor radiation must never drop below 97% which is a big challenge. Estimate for the RECEIVED

Analysis of the divertor ELM electron temperature at a uniquely high temporal resolution (1e-5 s) was reported at the JET tokamak [C. Guillemaut et al. Nucl. Fusion 58 (2018) 066006]. By collecting divertor probe data obtained during many dozens of ELMs, the conditional-average technique (CAV) yields surprisingly low peak electron temperatures, much below the pedestal ones (a reduction by 70% up to 99%!) which we, however, question. This result was interpreted through the collisional free-streaming kinetic model of ELMs, by a transfer of most of the electron energy to ions, implying a high tungsten sputtering for unmitigated ELMs in future fusion devices like ITER. Recently, direct microsecond temperature measurements on the COMPASS tokamak, however, showed that the electron temperature peak of ELM filaments measured in the divertor is reduced by less than ⅓ with respect to the pedestal one. This was further confirmed by a dedicated 1D PIC simulation and tends to prove that the pedestal electrons can transfer only their parallel energy to ions (due to low collisionality), thus less than ⅓ as is predicted by the collision-less free-streaming model. This finding is in strong contradiction with the JET observations. We have therefore compared the CAV to the direct (microsecond) ball-pen and Langmuir probes measurements in COMPASS and found very good agreement between them. Revisiting the mentioned JET CAV analysis yields indeed that the electron temperatures are much higher than previously reported, nearly as predicted by the PIC simulation, and thus the ion energy seems not significantly increasing in the SOL.

Three new in-vessel manipulators are designed and built for the new COMPASS Upgrade tokamak with uniquely high vessel temperature (250–500 °C) and heat flux density (perpendicular to divertor surface q ⊥ ∼ 80 MW/m2 and q ∥ ∼ GW/m2 at separatrix), which challenges the edge plasma diagnostics. Here we show their detailed engineering designs supported by heat conduction and mechanical models. Deep reciprocation of electrostatic probes near the separatrix should be possible by optimizing older concepts in (a) the head and probe geometry, (b) strongly increasing the deceleration up to 100× gravity by springs and strengthening the manipulator mechanical structure. One reciprocates close to the region of edge plasma influx (the outer midplane), the other at the plasma sink (between the outer divertor strike point and X-point), for studying the plasma divertor (impurity-seeded) detachment and liquid metal vapor transport. Both probe heads are equipped with a set of ball-pen and Langmuir probes, measuring reliably and extremely fast (10−6 s) local (1 mm resolution) plasma potential, density, electron temperature and heat flux and even ion temperature with 10−5 s resolution. The divertor manipulator (without reciprocation) will place various material test targets at the outer divertor. Unique will be its capability to increase 15× the surface heat flux with respect to the surrounding tungsten tiles just by controllable surface inclination of the test targets. We plan to test liquid metal targets where such inclined surface was found critical to achieve the desired mode with lithium vapor shielding. Even in the conservative expected performance of COMPASS Upgrade, we predict to reach and survive the EU DEMO relevant heat fluxes.

COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to main fields of COMPASS research obtained in the recent two years, including studies of turbulence, L-H transition, plasma material interaction, runaway electron, and disruption physics:  Tomographic reconstruction of the edge/SOL turbulence observed by a fast visible camera allowed to visualize turbulent structures without perturbing the plasma. Dependence of the power threshold on the X-point height was studied and related role of radial electric field in the edge/SOL plasma was identified. The effect of high-field-side error fields on the L-H transition was investigated in order to assess the influence of the central solenoid misalignment and the possibility to compensate these error fields by low-field-side coils. Results of fast measurements of electron temperature during ELMs show the ELM peak values at the divertor are around 80% of the initial temperature at the pedestal. Liquid metals were used for the first time as plasma facing material in ELMy H-mode in the tokamak divertor. Good power handling capability was observed for heat fluxes up to 12 MW/m 2 and no direct droplet ejection was observed.  Partial detachment regime was achieved by impurity seeding in the divertor. The evolution of the heat flux footprint at the outer target was studied.  Runaway electrons were studied using new unique systems -impact calorimetry, carbon pellet injection technique, wide variety of magnetic perturbations. Radial feedback control was imposed on the beam.  Forces during plasma disruptions were monitored by a number of new diagnostics for vacuum vessel motion in order to contribute to the scaling laws of sideways disruption forces for ITER. Current flows towards the divertor tiles, incl. possible short-circuiting through PFCs, were investigated during the VDE experiments. The results support ATEC model and improve understanding of disruption loads.