ELM-free H-mode Research Articles

Divertor materials evaluation system (DiMES)

The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4–18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Deuterium retention of different materials was measured using the 3He(d,p) 4He nuclear reaction (W.R. Wampler, R. Bastasz, D. Buchenauer et al., J. Nucl. Mater. 233 (1995) 791–797). For carbon, these measurements showed peak deuterium areal density of about 8 × 10 18 D/cm 2 in a co-deposited layer about 6 μm deep, mainly at the usually detached inboard divertor leg. That layer of carbon near the inner divertor strike point has an atomic saturation concentration of D/C ≈ 0.25, which is not significantly lower than the laboratory-measured saturation retention of 0.4. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 nm/s and the lowest deuterium uptake of 2 × 10 20 m 2.

ELMs and fast edge fluctuations in Alcator C-Mod

Just above the H-mode threshold power at low density, small amplitude high repetition rate (1-5 kHz) type III edge-localized modes (ELMs) are observed. At higher input power, the H-modes become ELM-free. At high density and high input power (, MW), the plasma often enters a new H-mode regime now called the enhanced H-mode, which is a steady-state H-mode often without distinct ELMs but with high levels of emission. Enhanced H-modes are accompanied by persistent high-frequency broadband (50-400 kHz) density and magnetic fluctuations. Above about , the enhanced H-modes may also exhibit bursting or continuous low m, n coherent magnetic fluctuations with frequencies from 60-90 kHz. In the H-mode, steep edge gradients are observed with a nine-channel electron cyclotron emission (ECE) polychromator and a 38-channel, high-resolution, edge soft-x-ray array. The soft-x-ray pedestal width remains around 3 to 4 mm in the ELM-free H-mode and between type III ELMs. The pedestal collapses within during an ELM. Calculations indicate that the plasma in the H-mode is often near ideal ballooning first stability limits, but no type I ELMs have yet been identified. Edge micro-tearing modes destabilized by a steep temperature gradient are also considered as a possible explanation of the observed fluctuations during the enhanced H-mode.

Transport properties and predictive modelling of JET H-mode and optimized shear discharges

An empirical model for anomalous transport in tokamak, supplemented by the proper boundary conditions, has been developed and applied to four different types of JET discharges: ELM-free hot-ion H-mode, L-mode, ELMy H-mode and discharges with optimized magnetic shear. The results of the numerical modelling show that the model gives quite reasonable agreement with experiment in a wide range of plasma parameters and operational regimes. Moreover, they reveal some important generic features of plasma transport which are model independent.

The influence of plasma-edge properties on high confinement discharges with a radiating plasma mantle at the tokamak TEXTOR-94

The radiative improved mode obtained on the limiter tokamak TEXTOR-94 combines the possibility of power exhaust by a radiating plasma boundary (with a fraction of the radiated power with respect to the total input power up to 90% with neon or argon cooling) with improved energy confinement (as good as in the ELM-free H-mode in divertor tokamaks) at high plasma densities (line-averaged central-electron density equal to or even above the Greenwald density limit ) in quasi-stationary discharges. An overview is given of the substantial changes in plasma-edge properties occurring at high radiated power levels . These changes are characterized by a reduction of the plasma-edge density and temperature, a reduction of particle transport out of the confined plasma volume and an increase of the penetration depth of deuterium and impurity atoms. As a consequence, the particle confinement time increases and the electron-density profiles steepen. The transition to improved confinement takes place as soon as the density peaking reaches a critical threshold. An internal transport barrier is observed in the bulk of RI-mode plasmas (at ) characterized by an increase of the pressure gradient and of the shear of the toroidal velocity compared to discharges without additional impurity seeding. The dilution at the plasma boundary is strongly increased by the seeded impurities whereas the central dilution is only weakly affected.

Latest JET results in deuterium and deuterium - tritium plasmas

All major JET systems have been fully commissioned for D - T and the DTE1 series of experiments has started with the D - T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D - D and D - T. In D-D, the L - H threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D - T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started. In the area of fusion performance, record D - D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D - T, internal transport barriers have been readily established in optimized shear discharges and Alfvén eigenmodes have been observed.

H-mode confinement and fusion performance in JET

JET has conducted extensive experiments in H-mode plasmas focusing on two broad aims. The first, involving the characterization of reactor-relevant aspects of the regime, encompasses the study of steady-state H-modes at plasma currents of up to 6 MA. In these plasmas, energy confinement is typically a factor of 1.8 - 2 greater than predicted by L-mode scalings and local transport varies with the ion Larmor radius, at least for input powers well above the H-mode threshold power. Using nitrogen impurity seeding, a regime has been developed in which good core confinement, with an enhancement of 1.5 relative to L-mode scaling, is combined with radiated power fractions of up to 80%. The second major goal is to optimize fusion power production by exploiting high-confinement ELM-free H-modes, which exhibit energy confinement enhancement factors of three or more relative to L-mode plasmas. In this regime, fusion power quotients have been achieved, though the highest fusion performance is transient.

Radiofrequency-heated enhanced confinement modes in the Alcator C-Mod tokamak

Enhanced confinement modes up to a toroidal field of BT=8 T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta (βN=1.5), low radiated power fraction (Pradmain/Ploss=0.3), and low effective charge (Zeff=1.5) has been obtained in Enhanced Dα H-mode. This type of H-mode has enhanced levels of continuous Dα emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q0.

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Helium transport and exhaust experiments in tokamaks

This is a review of recent experiments on helium transport and exhaust in tokamaks. The experiments are discussed in the light of requirements placed on alpha particle confinement for sustained ignition. The ratio of alpha particle replacement time to energy confinement time must be sufficiently small, and is used as a figure-of-merit for assessing the performance of experiments. The attainment of adequate helium removal depends most sensitively on recycling processes, pumping and boundary screening. Since no experiments exactly modeling the geometry or conditions in next step reactors have been performed, heavy emphasis is placed on identifying and understanding fundamental processes so that valid extrapolation is possible. Experiments have thus been carried out for basic regimes (L- and ELMy H-mode confinement, conventional high recycling divertor) and for improved performance conditions, both for the core (ELM-free H-mode, VH-mode, supershot, reverse shear, high β pol mode) and in the edge/divertor (completely detached H-mode, detached limiter). The core helium transport database has been accumulated mainly by gas puffing experiments, where the helium source is localized at the edge. Experiments with a central helium source from helium neutral beam injection find similar behavior, for the basic confinement regimes. Work is underway to make this comparison for improved confinement regimes. Results of integrated helium transport and exhaust experiments generally show adequate helium exhaust for next step reactors, but a widely used technique for helium exhaust (argon frost pumping) has been found to have important limitations for exploration of improved divertor (gas target) regimes. There has been less activity in the study of improved core and divertor operational modes, but a database is beginning to accumulate. While the figure-of-merit parameters for helium removal are favorable for next step reactors predicated on the basic core and divertor modes, the lack of similarity with next step devices requires the use of validated models even in this case, and these have been improved considerably.

Effects of edge and central fuelling on particle confinement in JT-60U

In order to understand the effects of the particle source distribution on the global particle confinement, the particle balance in JT-60U has been quantitatively analyzed. The dependence of the particle confinement time on the ratio of the central fuelling due to NBI to the edge fuelling due to wall recycling and gas-puffing was investigated. It was found that the particle confinement time increased with this ratio for the L-mode and ELMy H-mode plasmas, while, for the ELM-free H-mode plasma, the dependence of the particle confinement time on this ratio was not obvious. Furthermore, the confinement times of the center and edge fuelled particles were separated for the L-mode plasma. It was also found that the confinement time of the center fuelled particles was about 3 times longer than that of the edge fuelled particles.

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Profiles of T e and J sat measured by Langmuir probes at the target plates of the JET MkI divertor and the target plate power loading are modelled via the EDGE2D/NIMBUS codes. The low-density ( n e ∼ 1 × 10 19 m −3 ) OH plasma preceding high-power NBI, as well as the ELM-free high performance H ∗ and high performance rollover H RO phases are considered. Experimentally, the power splitting between ions ( P i ≥ 7 MW) and electrons ( P e < 1 MW) in conjunction with the H-mode transport barrier effects that T e and J sat change little from OH to H ∗ conditions, although P in ∼ 2 → 20 MW and n e can increase up to a factor of three as the H ∗-phase progresses. In modelling χ i is taken as 1 m 2/s. To match the divertor T e - profiles, χ e must be varied from the strike point outwards as ∼ 0.5 → 3 m 2/ s in all phases. Perpendicular particle transport is assumed to be purely diffusive or with a pinch. Duplication of J sat ( R) requires: D ⊥ ∼ 0.03 → 0.015 m 2/s (OH → H ∗) , or ν pinch D ⊥ ∼ 15 → 45 → 25 m −1 (OH → H ∗ → H RO) with D ⊥ = 0.1 m 2/ s. Sensitivity studies involving different recycling scenarios, deep and shallow computational grids, wall material and the thermal transport barrier are carried out to judge the possible influences on the deduced transport coefficients.

Effect of ELMS on the SOL plasma in DIII-D

We have studied the evolution of the edge plasma in VH-mode discharges in DIII-D, by following the changes in the radial profiles of the density and temperature, in the core plasma near the separatrix and in the scrape-off-layer (SOL) plasma outside the separatrix. The electron density and temperature profiles in the SOL do not show any significant difference between the ELM-free H-mode and VH-mode phases of the discharge. In the ELMing phase, the density profile broadens during an ELM, forming a high-density plateau that extends out into the SOL to the limit of the measurement. This plateau persists between the ELMs, although the density in the SOL does relax somewhat between the ELMs, with a characteristic time that can be greater than 10 ms, much longer than the sonic particle flow time to the divertor plates. The density scale length increases with the ELM background, as measured by the photodiode viewing the divertor floor nearest to, but outside, the outer strike point. The electron temperature profile in the SOL also broadens during an ELM, but the broad profile does not persist between ELMs.

H-modes on Alcator C-Mod

H-modes exhibiting improved confinement above the L-mode are achieved in Alcator C-Mod with ICRF and with ohmic heating alone without boronization. Both ELM-free and ELMy H-modes are obtained with total input power from 0.75 to 4.2 MW over a range of densities (0.8 to ) and toroidal fields (3 to 8 T). Type III ELMs are often observed to have coherent, high m and n precursor oscillations with frequencies of 100 - 160 kHz. The threshold power required to achieve the H-mode increases with density and toroidal field, in rough agreement with scalings derived from other tokamaks. The power densities and density times toroidal field products are an order of magnitude larger than in other tokamaks, in the range of values expected for ITER. The L - H and H - L transitions occur at approximately the same edge electron temperature. A low density limit to the H-mode is found at about . A high midplane neutral pressure limit of about 0.6 mTorr is also observed.

The H-mode in COMPASS-D

H-mode confinement is observed in COMPASS-D in ohmic and ECR heated single null divertor configuration discharges over a wide parameter range. At low densities the L - H transition power threshold shows a sharp increase, in contrast to scaling law behaviour. Transition to ELM-free H-mode is better correlated with a critical value of plasma pressure than edge power density. In experiments influencing H-mode dynamics, edge current density is found to be destabilizing, suggesting the possible role of the peeling mode. ELM characteristics are discussed, and impurity rotation observations and derived radial electric fields presented.

Ohmic H-modes in the TCV tokamak

The TCV tokamak has obtained ohmic H-modes in virtually all diverted plasmas with the ion drift directed towards an X-point and in several elongated limiter plasmas. Troyon factors up to 2 and line average densities up to , corresponding to Ithe Greenwald limit, have been obtained in diverted ELM-free H-modes. Quasi-stationary H-modes lasting for the entire current flat top (1.5 s) have been obtained in the presence of regular ELMs. The occurrence and magnitude of ELMs have been found to depend on configurational parameters such as the position of the `active' X-point in unbalanced double-null discharges and the plasma - wall separation in single-null discharges. These dependencies have permitted active control of ELM behaviour in TCV. A continuous spectrum of ELM amplitudes and frequencies has been observed, ranging from clearly identifiable type III ELMs to large, low-frequency ELMs which expel up to 12% of the stored energy and up to 7% of the particle content and are reminiscent of type I ELMs. A previously unknown, benign kind of ELM, with a maximum amplitude in the divertor region, has also been observed.

JET results with the new pumped divertor and implications for ITER

This paper presents an overview of results of the 1994/95 experimental campaign on JET with the new pumped divertor and draws implications for ITER in the areas of detached and radiative divertor plasmas, the use of beryllium as a divertor target tile material, the confinement properties of discharges with the same dimensionless parameters (except for the dimensionless Larmor radius) as ITER and the effect of varying the toroidal magnetic field ripple in the ITER relevant range. Discharges with high fusion performance at high current, in steady-state with ELMs and in the ELM-free hot-ion H-mode are also reported. Limits to operations are discussed and projections to D-T performance are made.

Transport modeling of core fueling and global particle confinement in ELM-free H-mode plasma in DIII-D

Evaluations of core fueling rates and global particle confinement times are presented for a single-null ELM-free H-mode discharge in the DIII-D tokamak. The edge plasma transport is modeled with the B2.5 code, and the neutral transport with both the DEGAS code and the internal neutrals model in the B2.5 code. Radial transport coefficients and corresponding plasma particle and heat fluxes into the scrape-off layer (SOL) are determined from power balance and by fitting the electron density and temperature profiles measured by Thomson scattering and the ion temperature profiles from charge-exchange/recombination spectroscopy. Divertor plasma parameters are deduced by fitting infrared television camera (IRTV) data, divertor Langmuir probe measurements, and D α data. Core fueling rates are calculated with DEGAS and compared to those required by particle balance, with the core efflux determined by B2.5. Global particle confinement times, deduced at four time slices during the 1.8 s ELM-free period, are typically 3–4 times the corresponding measured energy confinement times.

Ohmic H-mode power thresholds and recycling on COMPASS-D

Single null X-point discharges on COMPASS-D show a gradual L to H transition. Frequent boronization gives long ELM-free H-mode periods, duration > 100 ms, with continuously rising density to > 1.5 × 10 10 m −3. The LH transition threshold behaviour is more complex than that predicted by present simple scaling laws, with a well-defined low-density limit being observed and sensitivity to small perturbations at the plasma boundary. Density and recycling behaviour is analysed. In discharges refuelled by recycling alone, the coefficient of recycling as well as the particle confinement time is higher in H-mode periods.

Ohmic H-mode studies in TUMAN-3

The spontaneous transition from Ohmically heated limiter discharges into the regime with improved confinement termed as "Ohmic H-mode" has been investigated in "TUMAN-3". The typical signatures of H-mode in tokamaks with powerful auxiliary heating have been observed: sharp drop of Dalpha radiation with simultaneous increase in the electron density and stored energy, suppression of the density fluctuations and establishing the steep gradient near the periphery. The crucial role of the radial electric field in the L-H transition was found in the experiments with boundary biasing. The possibility of initiating the H-mode using single pellet injection was demonstrated. In Ohmic H-mode strong dependencies of tau E on plasma current and on input power and weak dependence on density were found. Thermal energy confinement time enhanced by a factor of 10 compared to predictions of Neo-Alcator scaling. Longest energy confinement time (30 ms) was obtained in the small tokamak TUMAN-3. Absolute values of the energy confinement time are in agreement with scaling proposed for description of the ELM-free H-modes in devices with powerful auxiliary heating ("DIII-D/JET H-mode" scaling).

Effects of ergodization on plasma confinement in JFT-2M

A steady-state H-mode plasma with frequent ELMs (edge localized modes) was obtained by applying magnetic fields with high poloidal mode numbers resonant in the plasma edge. The induced ELMs limit the density and impurity accumulation normally observed during ELM-free H-mode. An EML coil with numerous resonant modes (m10) in the plasma edge appears the most effective. This method of control is extended to higher auxiliary heating power by increasing the magnitude of the magnetic perturbations.