ELMing H-mode Plasmas Research Articles

Investigation of He–W interactions using DiMES on DIII-D

Tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ∼150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m−2.

Balance of the stored energies sustained by the internal and edge transport barriers and effects of ELMs and L–H transitions in JT-60U

To understand key physics processes determining radial profiles of the kinetic plasma parameters in the advanced tokamak operation scenarios, correlations between the edge transport barrier (ETB) and the internal transport barrier (ITB) have been studied in the JT-60U tokamak device. It has been found that the edge pedestal poloidal beta, βp-ped, increases almost linearly with the total poloidal beta, βp-tot, over a wide range of the plasma current for type I ELMing H-mode plasmas, and this dependence becomes stronger with increasing triangularity. This dependence is not due to the profile stiffness, since the dependence is the same regardless of the existence of ITB. As the stored energy inside the ITB-foot radius (WITB) increases, the total thermal stored energy (Wth) increases and then the pedestal stored energy (Wped) increases. On the other hand, as Wped increases, the ELM penetration expands more inwards and finally reaches the ITB-foot radius. At this situation, the ITB-foot radius cannot move outwards because of the erosion by ELMs. Then the fractions of WITB/Wth and Wped/Wth become almost constant. It has also been found that the type I ELM expels/decreases the edge toroidal momentum larger than the edge ion thermal energy. The ELM penetration for the toroidal rotation tends to be deeper than that for the ion temperature and can exceed the ITB-foot radius. The ELM penetration is deeper for CO-rotating plasmas than CTR rotating plasmas. In both cases, the ELM penetration is deeper in the order of the toroidal rotation (Vt), the ion temperature (Ti) and then the electron temperature (Te). The L–H transition also changes the Vt profile more significantly than the Ti profile. At the L–H transition, the pedestal Vt shifts into the CTR-direction deeply and suddenly without a change in Ti, and then the pedestal Vt grows further together with a growth of the pedestal Ti in a slower timescale. Such changes in Vt by ELMs and L–H transitions may affect degradation/evolution of ITBs.

Methane penetration in DIII-D ELMing H-mode plasmas

Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, 〈 n e〉 increases by <10%, energy confinement time, τ E, drops by <10%). The total carbon content is derived from C 6+ density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion ∇ B drift direction into the X-point, increasing the line averaged density from 5 to 8×10 19 m −3 dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (<1%). Changing the ion ∇ B drift to away from the X-point decreased the penteration factor by more than a factor of five for main chamber puffing.

Effect of plasma shape on particle flux and particle removal in DIII-D

Measurements from balanced or slightly unbalanced, double-null, ELMing H-mode plasmas show that the upper, more closed, divertor strike point particle fluxes are larger than the lower, more open, divertor strike point particle fluxes for comparable conditions. Comparison of the upper and lower outer strike point peak particle fluxes measured with Langmuir probes during scans of the magnetic balance shows an asymmetry around the magnetic balance point which is consistent with both the enhanced upper recycling and the expected direction of private region E× B drifts. Comparison of the inner and outer peak particle fluxes in the upper divertor shows a dependence on magnetic field direction that is also consistent with the private region E× B drift direction.

Feasibility Study of a Compact Ignition Tokamak Based upon GyroBohm Scaling Physics

The design of a reduced size (R = 4.45 m, BT = 5.04 T) ignition tokamak (Q = ∞) with superconducting coils using a standard ELMing H-mode plasma appears to be feasible. This effective size (BT2/3R5/6) is smaller than current proposals for Q = 10 burning (D-T) plasma experiments. The good confinement required for ignition with this small effective size is obtained by operating along a gyroBohm scaling path starting from the existing tokamak database at high beta (β = 4.1%) so that the loss power from core transport exceeds the H-mode threshold power. Using a design that can achieve a high normalized current (Ip/aBT = 1.63) also helps to decrease the size of the machine. The design of this relatively compact ignition tokamak satisfies reasonable engineering constraints on the superconducting toroidal field coils and central solenoid, and allows for a sufficiently long burn time for the plasma current to relax to its final state.

The role of neutrals in the H–L back transition of high density single-null and double-null gas-fueled discharges in DIII-D

The role of neutrals in triggering the H–L back transition in high density ELMing H-mode plasmas is explored in double-null (DN) and single-null (SN) divertors. We propose that the neutral particle buildup below the X-point may play an important role in triggering the H–L transition at high density.

Gyroradius Scaling of Helium Transport

The scaling of the transport rate of helium ash with normalized gyroradius has been measured for the first time, utilizing ELMing H-mode plasmas on the DIII-D tokamak. The helium diffusivity is found to scale in a gyro-Bohm-like manner, similar to the thermal diffusivity. Even though the extrapolation in normalized gyroradius from DIII-D to a reactor-grade device is large, sensitivity studies indicate that helium ash dilution in such a device will be primarily dependent on the helium exhaust efficiency at the plasma edge and not on the core transport rates.

Comprehensive 2D measurements of radiative divertor plasmas in DIII-D

This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (partially detached divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2D ( R, Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. D α emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D 2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. L α and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.

Power balance in DIII-D during single-null ELMing H-mode plasmas

We account for ≥85% of the injected power in DIII-D during single-null diverted ELMing H-mode discharges. Core plasma radiation accounts for ≤18% with the rest flowing through the scrape-off layer to the inboard and outboard divertors. The total power roughly splits in an in: out ratio of 1:1.3. Of the power entering the outboard divertor 75% leaves as direct heat flux to the divertor target plates with the rest in the form of radiation. In the inboard divertor, however, 75% of the power leaves as radiation with the rest as divertor target plate heat flux and charge exchange losses. The in: out ratio remains constant over a range of injected power and plasma current and even as divertor conditions change due to gas puffing.

Helium transport and exhaust studies of H-mode discharges in the DIII-D tokamak

A collaborative program has been initiated on DIII-D to study helium (He) transport and exhaust in H-mode plasmas. To simulate the presence of He ash in DIII-D, a 50 ms He puff is injected into a DIII-D plasma, resulting in a He concentration of approximately 15%. The time dependence of the He density profiles in the plasma core is measured with charge-exchange recombination spectroscopy and the He spatial distribution on the divertor floor is studied with an impurity monitor array. The dependence of core transport diffusivities as a function of edge-localized modes (ELM) frequency has been studied and the first demonstration made of He exhaust from an ELMing H-mode plasma in a diverted tokamak. A measured value of tau *He/ tau E of 14 was obtained, which is within the acceptable range for a fusion reactor.

Helium transport in enhanced confinement regimes on the TEXTOR and DIII-D tokamaks

Comparisons of helium (He) transport and exhaust in L-mode and in an enhanced confinement regime (H-mode), which is induced by a polarizing electrode, have been made for the TEXTOR tokamak. The results show an increased tendency for He accumulation when bulk plasma energy and particle confinement are improved during the polarization induced H-mode. Since these results imply that a high He pumping efficiency may be necessary for H-mode burning plasmas, we have begun exploring He transport in a divertor H-mode, similar to that proposed for International Thermonuclear Experimental Reactor (ITER). A collaborative program has been initiated to measure He transport and scaling on DIII-D during L-mode, H-mode, and ELMing H-mode plasma conditions. To simulate the presence of He ash in DIII-D, a 25 ms He puff is injected into a DIII-D plasma resulting in a He concentration of {approx}5%. The time dependence of the He{sup 2+} density profiles in the plasma core is measured by charge-exchange recombination spectroscopy at 11 radial locations.

Effect of divertor bias on plasma flow in the DIII-D scrape-off layer

A toroidal ring bias electrode in contact with the single-null divertor scrape-off layer in the DIII-D tokamak achieved electric field control of plasma flow in the layer. Bias strongly altered the static pressure of neutralized gas in the new divertor chamber (designed for future pumping) during Ohmic, L-mode and ELMing H-mode plasma operation. The pressure increased or decreased, depending on the signs of the applied electric potential and toroidal magnetic field, in agreement with the E*B drift direction. The variation of the core plasma density was opposite to the variation of the pump chamber gas pressure and was proportionally much smaller. A large bias induced pressure rise appeared even when the scrape-off layer was not directed at the pump entrance aperture-a surprising effect that is interpreted qualitatively in terms of radial potential discontinuities. Bias also affected divertor heat fluxes, and preliminary observations are discussed qualitatively