Table of contents

An exact analytical solution of the equation of radiative transfer is obtained for a cylindrical system with specularly reflecting walls, accounting for the presence of polarization scrambling in the reflection process. The effects of polarization scrambling on the specific intensity of the radiation can be described, in part, via an effective wall reflection coefficient for the extraordinary (x) and ordinary (o) mode. For the special case of electron cyclotron radiation in a fusion plasma approaching reactor conditions, a numerical analysis of the impact of polarization scrambling on both the specific intensity of the radiation and the radial profile of the net power radiated as well as on the total power loss is carried out for ITER-like parameters in steady-state operation.

A new approach has been proposed to reconstruct the current density profile in tokamak plasmas. The boundary-only integral equation derived from the Grad–Shafranov equation, under the assumption of polynomial expansion of current density, will have no unknowns except for the polynomial expansion coefficients, once the magnetic flux and its derivative have been given along the plasma boundary with the aid of Kurihara's Cauchy-condition surface method based on magnetic sensor data. In addition to the discretized form of the equation, some constraints are taken into account: the total plasma current, zero-current along the plasma boundary and a scalar relationship derived from the MHD equilibrium to relate the current density to the magnetic flux. It is also assumed that the poloidal field, as a quantity closely related to the current density, can be measured at a certain number of points inside the plasma. The whole set of linear equations is solved using the singular value decomposition technique to determine the polynomial expansion coefficients. The validity of the present technique and the quality of the current density solution have been investigated through test calculations for some plasma configurations.

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, , is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, ; and by helium beam injection and measurement of the helium replacement time, . In ELMy H-mode, with plasma configuration optimized for pumping, is achieved, where is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be R_eff ≈ 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

Field reversed configurations (FRCs) have been formed and sustained for up to 50 normal flux decay times by rotating magnetic fields (RMFs) in the translation, confinement, and sustainment experiment. For these longer pulse times a new phenomenon has been observed: switching to a higher performance mode delineated by shallower RMF penetration, higher ratios of generated poloidal to RMF drive field, and lower overall plasma resistivity. This mode switching is always accompanied by, and perhaps triggered by, the spontaneous development of a toroidal field with a magnitude up to 20% of the peak poloidal field. The global data cannot be explained by previous RMF theory based on uniform electron rotational velocities or by numerical calculations based on uniform plasma resistivity, but agrees in many respects with new calculations made using strongly varying resistivity profiles. In order to more realistically model RMF driven FRCs with such non-uniform resistivity profiles, a double rigid rotor model has been developed with separate inner and outer electron rotational velocities and resistivities. The results of this modelling suggest that the RMF drive results in very high resistivity in a narrow edge layer, and that the higher performance mode is characterized by a sharp reduction in resistivity over the bulk of the FRC.

Methods are developed for evaluation of the drift orbit averaged particle source, ⟨S⟩, in realistic neutral beam injection scenarios. Finite Larmor radius (FLR) and finite orbit width effects are included providing an accurate treatment of beam injection geometries in both standard and spherical tokamaks. With the knowledge of ⟨S⟩ the full beam distribution is readily obtained using well-known Fokker–Planck modelling schemes in constant of motion variables. For the considered ITER beam injection scenarios the analysis provides distributions which are in good agreement with equivalent distributions calculated by the TRANSP analysis code. The inclusion of FLR effects is important for fusion devices with relatively small values of the magnetic field; this is illustrated by considering the beam injection scenario of a spherical tokamak.

Collective Thomson scattering (CTS) has been proposed as a viable diagnostic for characterizing fusion born α-distributions in ITER. However, the velocities of the planned 1 MeV deuterium heating beam ions in ITER are similar to that of fusion born α-particles and may therefore mask the measurements of the fusion products. We apply a new technique for calculating the orbit averaged source, ⟨S⟩, of beam ions for various ITER scenarios. With the known ⟨S⟩ Fokker–Planck modelling is applied to characterize the beam ions during the slowing down process. Theoretical CTS signals for both beam ions and the α-particles are calculated. Our investigations show that the CTS measurements of α-particles will not be masked by the presence of the beam ions in H-mode plasmas. In lower density reversed shear plasmas, only a part of the CTS α-particle spectrum will be perturbed.

A new MHD mode has been discovered during type-I ELMy H-modes in the Joint European Torus (JET) tokamak. This mode is excited by the perturbation of the edge plasma due to the edge localized mode (ELM). It is radially and poloidally well localized and has the toroidal mode number n = 1 and the poloidal mode number m = 3. The mode appears in different plasma configurations and in a wide range of global plasma parameters as long as the rational q = 3 surface is located in the ELM perturbed region. A possible explanation for the new mode is as the remnant of a magnetic island created by edge ergodization during the ELM. Consequences for the understanding of the ELM process itself are discussed.

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m²). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20–30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10⁻³ and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.