E. Barbato

Current drive at plasma densities required for thermonuclear reactors

Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confined in toroidal tokamak devices requires the development of efficient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing significant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

LHCD and coupling experiments with an ITER-like PAM launcher on the FTU tokamak

A prototype passive active multijunction (PAM) launcher for lower hybrid (LH) waves conceptually similar to that foreseen for ITER has been successfully tested on FTU at frequency f = 8 GHz. The power routinely and safely managed for the maximum time allowed by the LH power plant (0.9 s) without any fault in the transmission lines is 250 kW, corresponding to 75 MW m−2 across the antenna active area and very close to the design value of 270 kW or 80 MW m−2. The achieved value is at least 1.4 times larger than the ITER request, which would be only 52 MW m−2, if the 33 MW m−2 required to the ITER grill in order to couple 20 MW to the plasma, are scaled up linearly with f, from fITER = 5 GHz. This linear scaling of the power handling capability of the LH antennae is indeed conservative with respect to the available data. The test results validate also the other two main expectations relevant to ITER, foreseen by the codes, namely to operate with the grill entirely in the vessel shadow and to still preserve good current drive (CD) efficiency. Even with the grill mouth retracted 2 mm inside the port shadow and with density in front of the launcher very close or even lower than the cut-off value, the PAM reflection coefficient is always ≤ 2.5%, if the antenna has been properly conditioned. The CD efficiency is comparable to that of a conventional grill, once the lower directivity is taken into account. Flexibility in determining the N|| spectrum is also maintained, according to hard x-rays and electron cyclotron emission spectra. Conditioning the PAM in order to operate at the ITER equivalent power level required only one day of radio-frequency operation, without a previous baking of the waveguides.

Towards fully non-inductive current drive operation in JET

Quasi-steady operation has been achieved at JET in the high-confinement regime with internal transport barriers (ITBs). The ITB has been maintained up to 11 s. This duration, much larger than the energy confinement time, is already approaching a current resistive time. The high-performance phase is limited only by plant constraints. The radial profiles of the thermal electron and ion pressures have steep gradients typically at mid-plasma radius. A large fraction of non-inductive current (above 80%) is sustained throughout the high-performance phase with a poloidal beta exceeding unity. The safety factor profile plays an important role in sustaining the ITB characteristics. In this regime where the self-generated bootstrap current (up to 1.0 MA) represents 50% of the total current, the resistive evolution of the non-monotonic q-profile is slowed down by using off-axis lower-hybrid current drive.

Measurement of the hot electrical conductivity in the PBX-M tokamak

A new method for the analysis of tokamak discharges in which the plasma current is driven by a combination of high power RF waves and a DC electric field is presented. In such regimes, which are the most usual in RF current drive experiments, it is generally difficult to separate the different components of the plasma current, i.e. purely ohmic, purely non-inductive and cross terms. If the bilinear (in wave power and electric field) cross-term is the dominant one, an explicit relation between the loop voltage drop and the injected power can be found. This relation involves two parameters, the purely RF current drive efficiency and the hot (power dependent) electrical conductivity. These can be simultaneously determined from a simple two parameter fit, if the loop voltage drop is measured at several RF power levels. An application to lower hybrid current drive experiments in the Princeton Beta Experiment (PBX-M) tokamak is presented. It is shown that the method also allows independent evaluation of the average power absorption fraction and the n|| upshift

Integrated modelling of the current profile in steady-state and hybrid ITER scenarios

We present integrated modelling of steady-state and hybrid scenarios for ITER parameters using several predictive transport codes. These employ models for non-inductive current drive sources in conjunction with various theory-based and semi-empirical transport models. In conjunction with the simulation effort, the current drive models are being evaluated in a series of cross-code and code-experiment comparisons under ITER-relevant conditions. New benchmark evaluations of current drive from injection of neutral beams (NBCD), electron cyclotron waves (ECCD) and lower hybrid waves (LHCD) are reported. Simulations using several transport modelling codes self-consistently calculate the heating and current drive sources using ITER design parameters. Operating constraints are also taken into account, although the calculations reported here still require further refinement. The modelling addresses both the final stationary state and dynamic access to it. The simulations indicate that generation and control of internal and edge barriers to access and maintain high confinement will be a major undertaking for future simulations, as well as a challenge for the ITER steady-state and hybrid experimental programme.

Progress in LHCD : a tool for advanced regimes on ITER

The recent success in coupling lower hybrid (LH) waves in high performance plasmas at JET together with the first demonstration on FTU of the coupling capability of the new passive active multijunction launcher removed major concerns on the possibility of using LH on ITER. LH exhibits the highest experimental current drive (CD) efficiency at low plasma temperature thus making it the natural candidate for off-axis CD on ITER where current profile control will help in maintaining burning performance on a long-time scale. We review recent LH results: long internal transport barrier obtained in JET with current profile sustained and controlled by LH acting under real time feedback together with first LH control of flat q-profile in a hybrid regime with T e ∼ T i . Minutes long fully non-inductive LH driven discharges on Tore Supra (TS). High CD efficiency with electron cyclotron in synergy with LH obtained in FTU and TS opening the possibility of interesting scenarii on ITER for MHD stabilization. Preliminary results of LH modelling for ITER are also reported. A brief overview of ITER LH system is reported together with some indication of new coming LH experiments, in particular KSTAR where CW klystrons at the foreseen ITER frequency of 5 GHz are being developed.

Progress towards internal transport barriers at high plasma density sustained by pure electron heating and current drive in the FTU tokamak

Strong electron internal transport barriers (ITBs) are obtained in FTU by the combined injection of lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC, up to 0.8 MW) radio frequency waves. ITBs occur during either the current plateau or the ramp-up phase, and both in full and partial current drive (CD) regimes, up to peak densities ne0>1.2×1020 m−3, relevant to ITER operation. Central electron temperatures Te0>11 keV, at ne0≈0.8×1020 m−3 are sustained longer than 35 confinement times. The ITB extends over a region where a slightly reversed magnetic shear is established by off-axis LHCD and can be as wide as r/a = 0.5. The EC power, instead, is used either to benefit from this improved confinement by heating inside the ITB, or to enhance the peripheral LH power deposition and CD with off-axis resonance. Collisional ion heating is also observed, but thermal equilibrium with the electrons cannot be attained since the e−–i+ equipartition time is always 4–5 times longer than the energy confinement time. The transport analysis performed with both ASTRA and JETTO codes shows a very good relation between the foot of the barrier and the weak/reversed shear region, which in turn depends on the LH deposition profile. The Bohm-gyroBohm model accounts for the electron transport until Te0<6 keV, but is pessimistic at higher temperatures, where often also a reduction in the ion thermal conductivity is observed, provided any magnetohydrodynamic activity is suppressed.

Reduction of the electron thermal conductivity produced by ion Bernstein waves on the Frascati Tokamak Upgrade tokamak

Operating with a high frequency and a wave guide antenna, the ion Bernstein wave (IBW) experiment on the Frascati Tokamak Upgrade is not dominated, as expected, by nonlinear plasma edge phenomena. By coupling IBW power, a simultaneous increase of plasma density and central electron temperature (⩾2 keV) is produced when the confinement magnetic field is adjusted to set an ion cyclotron resonant layer in the plasma bulk. Transport analysis indicates a reduction of the electron thermal transport inside the internal resonant layer larger than a factor of 2.

Absorption of lower hybrid wave power by α-particles in ITER-FEAT scenarios

The data in figures 6 and 13(a) were incorrectly plotted. Please see the pdf for the corrected versions.

The role of non-resonant collision dissipation in lower hybrid current driven plasmas

In this paper the role of non-resonant collision absorption (NRCA) of lower hybrid (LH) wave power is analysed using a combined ray-tracing Fokker–Planck model. The analysis shows how much and under what conditions NRCA affects the current drive efficiency (CDE) and the effective ability of LH wave to penetrate into the plasma thus generating a fast electron tail. In all LH experiments the CDE is shown to increase as a function of the volume-averaged electron temperature up to a saturation level. Such dependence can be attributed to NRCA, at the plasma periphery, that turns out to be more relevant, the lower the plasma temperature. Furthermore, NRCA is shown to be a key issue in interpreting recent results on LHCD experiments on FTU, where a new regime is found of high-density discharges with a hot edge, where clear signs of fast electron tail generation are present. In standard high-density discharges with a cold edge conversely, LHCD effects are not observed. The numerical calculations reported here clearly show that the difference between these two regimes can be entirely attributed to the difference in NRCA at the plasma periphery. In conclusion, the paper shows that collision absorption can be responsible for the degradation of the CDE in low-temperature plasmas and can completely prevent the penetration of LHCD in high-density cold edge plasmas.