Table of contents

New experiments with the ion Bernstein waves (IBWs) have been performed in the Frascati Tokamak Upgrade (FTU) both in hydrogen and deuterium plasmas at higher power level, operation at higher density, higher plasma current and lower effective ion charge values than in the previous campaign. Improved confinement in a region with larger radius is obtained when operating in deuterium. Transport analysis shows a uniform decrease of the electron thermal conductivity by 40% over the radial region bounded by the absorption layer.

Feedback stabilization of magnetohydrodynamic modes in reversed field pinches is analysed for a set of discrete coils driven by voltage control. It is found that the resistive wall mode can be stabilized with a very simple controller structure and with acceptable voltages in the coils. These results are obtained by using a sufficient number of active coils and either sensors for the radial field or sensors for the poloidal or toroidal field placed inside the resistive wall. The result is robust with respect to variations in the plasma equilibrium. Poloidal and toroidal sensors placed outside the wall require a more complicated controller and very high voltages, and do not allow as good control performance as internal sensors.

Injection of frozen deuterium pellets from the magnetic low field side into discharges of the Frascati Tokamak Upgrade with high current and broad temperature profile gives rise to strong central fuelling, whereas pellet ablation calculations predict that pellets should be almost completely sublimated outside the q = 1 magnetic surface. Enhanced central fuelling has been observed in many tokamaks; in these experiments, strong temperature asymmetries are observed inside the q = 1 surface in addition to anomalous central fuelling. Both phenomena are explained by the rapid growth of an m = 1 magnetic island. Detailed simulations of temperature evolution have been performed by means of a modified version of the M1TEV code in which the density evolution has been added.

Power deposition profiles on JET MkIIGB divertor plates were measured for a variety of plasma conditions. Comparison of experimental measurements with theoretical predictions of two dozen candidate models of perpendicular heat conductivity suggests that a combination of classical ion conduction and ion orbit loss from the pedestal region dominates radial energy transport in the scrape-off layer (SOL), at least in the vicinity of the separatrix, during the inter-edge localized mode (ELM) phase of H-mode discharges. Although not a conclusive proof of suppression of (ion) turbulence, it is consistent with the hypothesis that the edge transport barrier extends into the SOL during the inter-ELM phase. Extrapolations to ITER predict an outer target power width of 3 ± 1 mm-omp (mapped to the outer mid-plane) at the entrance into the divertor volume.

Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side of the ³He ion cyclotron resonance layer in D and ⁴He plasmas and subsequently damped on the bulk electrons. The resulting electron power deposition, measured using ICRF power modulation, is narrow with a typical full-width at half-maximum of ≈30 cm (i.e. about 30% of the minor radius) and the total deposited power to electrons comprises at least up to 80% of the applied ICRF power. The ICRF mode conversion power deposition has been kept constant using ³He bleed throughout the ICRF phase with a typical duration of 4–6 s, i.e. 15–40 energy confinement times. Using waves propagating in the counter-current direction minimizes competing ion damping in the presence of co-injected deuterium beam ions.

Studies of global energy confinement and toroidal plasma current behaviour for the second harmonic 70 GHz ECH at B = 1–1.5 T are described with emphasis on the magnetic configuration effects in the helical-axis heliotron 'Heliotron J'. At low densities of , the electron temperature reached T_e ≈ 1 keV in the core region, indicating the production of collision-less plasmas of electron collisionality ν^* ≪ 0.1, where ν^* = ν/(v_e/π R₀q). For medium densities of , the preferable energy confinement time, 1.5–2 times larger than that of the ISS95 scaling, was obtained under the condition of localized central heating at B ≈ 1.25 T for the standard configuration of Heliotron J. The measurements of the toroidal current under perpendicular microwave injection revealed the change of the current flow direction as a function of the poloidal magnetic field. The measured current behaviour was found to be qualitatively consistent with that of the bootstrap current predicted from neoclassical theory. The observed flow reversal showed that a proper selection of the field configuration could control the bootstrap current in the helical-axis heliotron. In addition, the current control through the electron cyclotron current drive scenario with oblique injection of microwaves was experimentally examined.

Reduced activation ferritic/martensitic (RAFM) steels are the reference structural materials for future fusion reactors. They have proven to be a good alternative to austenitic steels for their higher swelling resistance, lower damage accumulation and improved thermal properties. However, irradiated RAFM steels exhibit a low temperature hardening and an increase in the ductile-to-brittle transition temperature, which imposes a severe restriction on reactor applications at temperatures below about 350°C. Furthermore, a high density of small cavities (voids or helium bubbles) has been recently evidenced in specimens irradiated with a mixed spectrum of neutrons and protons at about 300°C at a dose of 10 dpa, which could affect their fracture properties at intermediate temperatures. The upper temperature for the use of RAFM steels is presently limited by a drop in mechanical strength at about 500°C. New variants that can withstand higher temperatures are currently being developed, mainly using a stable oxide dispersion. This paper reviews European activity in the development of RAFM steels.

In order to study the effect of carbon impurities in plasmas on tungsten first walls, a hydrogen–carbon mixed ion beam was irradiated on tungsten. It was found that a very small amount of carbon (∼0.3% or more) significantly enhanced blister formation. Small blisters appeared in a fluence of ∼10²³ H m⁻² and their number increased in the 10²⁴ H m⁻² range. Then, the size of the blisters increased towards the 10²⁵ H m⁻² range. According to XPS analysis, all the tungsten atoms near the top surface were combined with carbon atoms to form a tungsten carbide layer in the cases where significant blisters were observed. This layer prevented the implanted H from leaving the tungsten. As a result, it caused enhanced diffusion into the bulk, leading to blister formation.

Neural networks are trained to evaluate the risk of plasma disruptions in a tokamak experiment using several diagnostic signals as inputs. A saliency analysis confirms the goodness of the chosen inputs, all of which contribute to the network performance. Tests that were carried out refer to data collected from succesfully terminated and disruption terminated pulses performed during two years of JET tokamak experiments. Results show the possibility of developing a neural network predictor that intervenes well in advance in order to avoid plasma disruption or mitigate its effects.

Feedback control of nonaxisymmetric resistive wall modes is studied analytically for cylindrical plasmas and computationally for high beta tokamaks. Internal poloidal sensors give superior performance to radial sensors, for instance in terms of the highest achievable plasma pressure. A single poloidal array of feedback coils allows robust control with respect to variations in plasma pressure, current and rotation velocity. The control analysis is applied to advanced scenarios for ITER. Configurations with multiple poloidal coils and feedback systems for nonresonant MHD instabilities in reversed field pinches are also studied. The control study was carried out using the assumption of ideal amplifiers.

The influence of dominant electron heating on internal transport barriers (ITBs) was investigated in JT-60U. The electron cyclotron waves were mainly utilized in order to extend the experimental space into a dominant electron heating regime. The experiments were carried out on both reversed magnetic shear (RS) and positive magnetic shear (PS) plasmas. It was found that in RS plasmas clear ITBs in T_e, T_i and n_e profiles remained even under dominant electron heating. Energy confinement in excess of twice the H-mode scaling has been obtained in the regime. On the other hand, it was found that in PS plasmas the T_i ITB was degraded by dominant electron heating.

Defects in the form of vacancies (loops, voids, etc) created by hydrogen implantation into selected Cu alloys foreseen for the International Thermonuclear Experimental Reactor (ITER) first wall cladding were studied using positron annihilation spectroscopy. The pulsed low-energy positron system, which enables depth profiling of the positron lifetime spectra in the near-surface region (20–460 nm) of hydrogen-implanted copper alloys, was applied, and its results were compared with TRIM calculations and transmission electron microscopy studies. The selected specimens were implanted in the Ion Beam Laboratory of FEI STU Bratislava. The energy of implantation was E_H = 2 × 95 keV for the molecular ion beam. The temperature during this process was lower than 90°C. Two implantation doses were chosen for both the alloys: 1.3 × 10¹⁹ ions cm⁻² (1.1 C cm⁻²) and 5 × 10¹⁸ ions cm⁻² (0.4 C cm⁻²). Although the influences of neutrons with energy 14 MeV and protons with energy 95 keV are not the same (differences in energy and existence of proton charge), experimental simulation of radiation damage of ITER construction materials was successfully performed. The results are discussed in terms of microstructural changes of the studied materials upon irradiation. The CuAl25 alloy seems to be more resistant to proton bombardment than CuCrZr.

The fast ignition concept of laser fusion utilizes hot electrons produced at the surface of the target by an incident intense laser pulse for the creation of the hot spot for ignition. As the hot electrons move inwards to the core of the precompressed target, the electrons from the background plasma provide a return shielding current. Three-dimensional PIC simulations have shown that intense Weibel, tearing and coalescence instabilities take place which organize the current distribution into a few current filaments. In each of these filaments the central core region constitutes a current due to the fast electrons propagating inwards towards the pellet core, while the outer cylindrical shell region carries the return shielding current. The presence of instabilities and their subsequent nonlinear development can hinder the propagation of fast electrons towards the core influencing the location of the hot spot for ignition. Earlier studies showing the existence of sausage-like modes were carried out in the non-relativistic limit and under the assumption of equal electron densities of the fast and the cold electrons. The fast electron density, in general, differs considerably from the background plasma density as it is dependent on the incident laser intensity. This paper incorporates relativistic effects and also studies the dependence of the growth rate on the fast electron density. Finally, nonlinear saturation of the instability and its impact on the stopping of the fast electron motion towards the core have also been investigated using numerical simulations. The simulations have, however, currently been carried out for non-relativistic dynamics. The results show that the sheared velocity profile of the channel gets flattened, causing an effective drop in the inward moving current.

This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β_N = 1.8, HH = 1.0, and fn_GW = 0.85 had sufficient potential to achieve the electric break-even condition (net electric power ) under the following engineering conditions: machine major radius 6.5 m ⩽ R_p ⩽ 8.5 m, the maximum magnetic field on TF coils B_tmax = 16 T, thermal efficiency η_e = 30%, and NBI system efficiency η_NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β_N and fn_GW. β_N ⩾ 3.0 is required for with fusion power P_f ∼ 3000 MW. On the other hand, fn_GW ⩾ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T_ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P_f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R_p ∼ 7.5 m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy.

The eigenfunctions of Rayleigh–Taylor modes on a spherical capsule are the spherical harmonics Y_l,m. These can be measured by measuring the surface perturbations along great circles and fitting them to the first few modes by a procedure described in this paper. For higher mode numbers, it is more convenient to average the Fourier power spectra along the great circles and then transform them to spherical harmonic modes by an algorithm derived here.

ELMy H-mode experiments at JET in 2000/mid-2002 have focused on discharges with normalized parameters for plasma density, energy confinement and beta similar to those of the ITER Q_DT = 10 reference regime (n/n_GW ∼ 0.85, H_98(y,2) ∼ 1, β_N ∼ 1.8). ELMy H-mode plasmas have been realized reaching or even exceeding those parameters in steady-state conditions (up to ∼ 5 s or 12 τ_E) in a reproducible way and only limited by the duration of the additional heating phase. These results have been obtained (a) in highly triangular plasmas, by increasing the average triangularity δ towards the ITER reference value (δ ∼ 0.5), and (b) in plasmas at low triangularity (δ ∼ 0.2) by seeding of Ar and placing the X-point of the plasma on the top of the septum. Pellet injection from the high field side is a third method yielding high density and high confinement, albeit not yet under steady-state conditions. In highly triangular plasmas the influence of input power, plasma triangularity and impurity seeding with noble gases has been studied. Density profile peaking at high densities has been obtained in (a) impurity seeded low triangularity discharges, (b) ELMy H-modes with low levels of input power and (c) discharges fuelled with pellet injection from the high field side. New ELM behaviour has been observed in high triangularity discharges at high density, opening a possible route to ELM heat load mitigation, which can be further amplified by Ar impurity seeding. Current extrapolations of the ELM heat load to ITER show possibly a window for Type I ELM operation. Confinement scaling studies indicate an increase in confinement with triangularity and density peaking, and a decrease in confinement with the Greenwald number. In addition, experiments in H isotope and He indicate τ_E ∝ M^0.19Z^−0.59. The threshold power for the L–H transition in He plasmas shows the same parametric dependence as in D plasmas, but with a 50% higher absolute value.

The two-dimensional radial vs poloidal structure and motion of edge turbulence in the National Spherical Torus Experiment (NSTX) were measured using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analysed for Ohmic, L- and H-mode plasma conditions. The two-dimensional images often show regions of strong localized light emission known as 'blobs', which move both poloidally and radially at a typical speed of ≈10⁵ cm s⁻¹, and sometimes show spatially periodic features.

The 'effective' fuelling efficiency of hydrogen gas puffing ranges from 10% to 50% in the Large Helical Device. A local increase in neutral particle pressure at the gas puff port was measured in the experiment. The pressure increase rate corresponds to ∼10% of the gas puff flux. The other 90% of the gas puff flux increases the density and/or the plasma outflow. A particle balance model reveals that the recycling flux estimated from the particle flux on the divertor plates increases during the gas puffing. It is shown that the high effective fuelling efficiency is possibly due to the large recycling flux. At the limit of small recycling flux, the effective fuelling efficiency decreases to ∼10%. In the helium gas puff discharge, the effective fuelling efficiency is larger than the hydrogen gas puffing and approaches 100%. This can be related to the large recycling coefficient of more than 0.95.

A non-linear Grad–Shafranov toroidal equilibrium reconstruction code (MSTFit) has been developed for the Madison Symmetric Torus. This is the first such code applied to the unique magnetohydrodynamic (MHD) equilibrium of the reversed field pinch. A new set of toroidal Green's tables have been computed to impose the boundary condition of the close-fitting conducting shell. The non-linear fitting routine is sufficiently versatile for incorporating data from a variety of internal and external diagnostics, including a novel constraint based on orbits from a heavy ion beam probe diagnostic. Utilizing the full complement of internal and external magnetic and pressure diagnostics, MSTFit resolves accurately subtle changes in internal magnetic structure with implications on MHD stability. We show example equilibria that confirm conservation of magnetic helicity during relaxation and two-dimensional equilibrium effects.

The aim of this paper is to report on recent advances made in global gyrokinetic simulations of ion temperature gradient (ITG) modes and other microinstabilities. The nonlinear development and saturation of ITG modes and the role of E × B zonal flows are studied with a global nonlinear δf formulation that retains parallel nonlinearity and thus allows for a check of the energy conservation property as a means of verifying the quality of the numerical simulation. Due to an optimized loading technique, the conservation property is satisfied with an unprecedented quality well into the nonlinear stage. The zonal component of the perturbation evolves to a quasi-steady state with regions of ITG suppression, strongly reduced radial energy flux and steepened effective temperature profiles alternating with regions of higher ITG mode amplitudes, larger radial energy flux and flattened effective temperature profiles. A semi-Lagrangian approach free of statistical noise is proposed as an alternative to the nonlinear δf formulation. An ASDEX-Upgrade experiment with an internal transport barrier is analysed with a global gyrokinetic code that includes trapped electron dynamics. The weakly destabilizing effect of trapped electron dynamics on ITG modes in an axisymmetric bumpy configuration modelling W7-X is shown in global linear simulations that retain the full electron dynamics. Finite β effects on microinstabilities are investigated with a linear global spectral electromagnetic gyrokinetic formulation. The radial global structure of electromagnetic modes shows a resonant behaviour with rational q values.

The toroidal field coils in Tore Supra are supra-conducting, and their number is restricted to 18. As a result, the ripple is fairly large, about 7% at the plasma boundary. Tore Supra has consequently been equipped with dedicated ripple loss diagnostics, which has allowed ripple loss studies. This paper reports on the measurements made with these diagnostics and provides an analysis of the experimental results, comparing them with theoretical expectations whenever possible. Furthermore, the main heating source accelerating ions in Tore Supra is ion cyclotron resonance range of frequency (ICRF) heating, and the paper provides new information on the ripple losses of ICRF accelerated ions.

Fusion research has focused on the goal of a fusion power source that utilizes deuterium–tritium (D–T) because the reaction rate is relatively large. Fusion reactors based on the deuterium–deuterium (D–D) reaction, however, might be superior to D–T based reactors insofar as they minimize the power produced in neutrons and do not require the breeding of tritium. We explore an alternative D–D based fuel cycle and show that a levitated dipole may be uniquely suited for this application. We find that a dipole based D–D power source can potentially provide a substantially better utilization of magnetic field energy with a mass power density comparable with a D–T based tokamak power source.

Pedestal studies in DIII-D find that the width of the region of steep gradient in the H-mode density is comparable with the neutral penetration length, as computed from a simple analytic model. This model has analytic solutions for the edge plasma and neutral density profiles, which are obtained from the coupled particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40 and 500 eV), the analytic model quantitatively predicts the observed decrease in the width as the pedestal density increases and the observed strong increase in the gradient of the density as the pedestal density increases. The model successfully predicts that L-mode and H-mode profiles with the same pedestal density have gradients that differ by less than a factor of 2. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fuelling is an important part of the physics that determines the structure of the H-mode transport barrier.