Table of contents

The effect of divertor baffling on impurity neutral and impurity ion compression in ohmic discharges in the Alcator C-Mod tokamak is explored. Experiments are performed using a novel divertor bypass which allows in situ variations to the divertor baffling. The results indicate that divertor baffling in Alcator C-Mod plays an important role in impurity neutral compression, but little role in impurity ion compression. The apparent separation between neutral compression and ion compression is surprising, but is consistent with a model in which the neutral compression (both in deuterium and argon) is determined by a leakage flux through the baffle structure from the divertor region to the main chamber which is independent of the leakage conductance. Instead, the leakage fluxes appear to be determined by the divertor plasma conditions (including impurity and fuel ions), which are not influenced by the baffle structure.

The effect of first wall reflectivity on synchrotron radiation loss in tokamak reactors of arbitrary geometry is discussed. A commonly employed approximation is revisited, in the light of a recent accurate scaling law for global synchrotron loss in tokamaks of arbitrary geometry. It is shown that the synchrotron loss expected in reactors with highly reflecting walls is reduced with respect to previous estimates.

The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. A straightforward extrapolation of Wendelstein 7-X leads to HSR5/22, which has 5 field periods and a major radius of 22 m. HSR4/18 is a more compact Helias reactor with 4 field periods and an 18 m major radius. Stability limit and energy confinement times are nearly the same as in HSR5/22, thus the same fusion power (3000 MW) is expected in both configurations. Neoclassical transport in HSR4/18 is very low, and the effective helical ripple is below 1%. The article describes the power balance of the Helias reactor, and the blanket and maintenance concepts. The coil system of HSR4/18 comprises 40 modular coils with NbTi superconducting cables. The reduction from 5 to 4 field periods and the concomitant reduction in size will also reduce the cost of the Helias reactor.

Several innovative applications of a travelling wave (combline) antenna designed for fast wave current drive have been demonstrated for the first time in the JFT-2M tokamak. High energy electrons of at least 10 keV were produced in the plasma core by highly directional fast waves in electron cyclotron heated plasmas. The ponderomotive potential of the beat wave, produced by fast waves at two different frequencies, was directly measured for the first time by a heavy ion beam probe. Plasma production was demonstrated using the wave fields excited by the combline antenna over a wide range of toroidal magnetic fields (0.5-2.2 T).

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at I_p = 0.8 MA and B_T = 1.6 T is comparable to or smaller than that for JT-60U IPP at I_p = 1.0 MA, B_T = 3.8 T and I_p = 1.5 MA, B_T = 3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated D_α emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region.

Edge conditions in DIII-D are being quantified in order to provide insight into the physics of the H mode regime. Several studies show that electron temperature is not the key parameter that controls the L-H transition. Gradients of edge temperature and pressure are much more promising candidates for elements of such parameters. They systematically increase during the L phases of discharges which make a transition to H mode, and these increases are typically larger than the increases in the underlying quantities. The quality of H mode confinement is strongly correlated with the height of the H mode pedestal for the pressure. The gradient of the pressure is limited by MHD modes, in particular by ideal kink ballooning modes with finite mode number n. For a wide variety of discharges, the width of the barrier for electron pressure is well described by a relationship that is proportional to (β^ped_p)^1/2. A new regime of confinement, called the quiescent H mode, which provides steady state operation with no ELMs, low radiated power and normal H mode confinement, has been discovered. A coherent edge MHD mode provides adequate particle transport to control the plasma density while permitting the pressure pedestal to remain almost identical to that observed in ELMing discharges.

Plasma confinement with a transport barrier such as an H mode or an internal transport barrier mode is examined under the constraint due to the conservation of total angular momentum. The results are tested against actual experimental data and the general characteristics of plasma confinement with a transport barrier are well understood in terms of this constraint. This implies that the confinement of tokamak plasmas can be determined by the decay rate of the total angular momentum. It also suggests that the confinement with a transport barrier is good, since electrostatic fluctuations cannot affect this constraint, but that electromagnetic fluctuations such as ELMs can cause the confinement to deteriorate.

High β_N conditions have been sustained in JET discharges with an internal transport barrier (ITB) by controlling edge conditions and pressure peaking. The behaviour of ideal and resistive MHD instabilities at high β_N has been studied as a function of pressure peaking. The duration of the high β_N phase was limited by interaction with the septum of the Gas Box divertor. ITB triggering and evolution are in good agreement with the condition for turbulence suppression by E × B shear flow.

High power ICRF heating of a hydrogen minority ion species in JET tritium plasmas has generated a total neutron rate that is about 40% larger than the 14 MeV neutron rate originating from fusion reactions between bulk tritium ions and deuterium minority ions. The T(p,n)³He fusion reaction, caused by ICRF accelerated protons, is identified as a source for producing the excess neutron emission. This reaction is endothermic and has a proton energy threshold of about 1 MeV and a peak cross-section at about 3.0 MeV. The presence of protons with such high energies is detected in gamma ray and high energy neutral particle analyser measurements and is also confirmed by ICRF modelling with the PION code. The fast proton energy content and the pT fusion reactivity as simulated by the PION code are compared with the experimental measurements when classical slowing down and confinement of ICRF accelerated protons are assumed in the simulations.

A new method of actively modifying the plasma-wall interaction was tested on the Tokamak Fusion Test Reactor. A laser was used to introduce a directed lithium aerosol into the discharge scrape-off layer. The lithium introduced in this fashion ablated and migrated preferentially to the limiter contact points. This allowed the plasma-wall interaction to be influenced in situ and in real time by external means. Significant improvement in energy confinement and fusion neutron production rate as well as a reduction in the plasma Z_eff have been documented in a neutral beam heated plasma. The introduction of a metallic aerosol into the plasma edge increased the internal inductance of the plasma column and also resulted in prompt heating of core electrons in ohmic plasmas. Preliminary evidence also suggests that the introduction of an aerosol leads to both edge poloidal velocity shear and edge electric field shear.

The edge fluctuations and transport in the HT-7 tokamak are investigated using a Langmuir probe array in ohmic and IBW heated discharges. The normalized fluctuation levels are in the range 15-60% and 5-20% for electron density and temperature, respectively, and have the non-Boltzmann relation e_f/T_e > ñ_e/n_e > _e/T_e in the SOL of the ohmically heated plasmas. The fluctuation spectra show typical broadband features as observed in other tokamaks. In IBW heated plasmas, the particle confinement is greatly improved and the poloidal velocity shear in the SOL is strongly modified in a manner that is equivalent to an additional poloidal velocity in the electron diamagnetic direction produced by the IBW. A decorrelation in the fluctuations and the suppression of turbulent transport by the effect of the poloidal shear are found to be common mechanisms that are observed from the radial profiles of the fluctuating quantities in all these discharges. Electrostatically driven turbulent transport can account for a significant part of global particle losses in both ohmic and IBW heated plasmas.

Utilization of electron cyclotron radiofrequency sources for current drive and the stabilization of neoclassical tearing modes (NTMs) in next step devices rests on the current density and current drive efficiency attainable. Optimization is reported of electron cyclotron driven current density and current drive efficiency with respect to source frequency as well as toroidal and poloidal launch angles for two launcher positions: an `upper port' position above the midplane and a `midplane' position at the height of the magnetic axis. The plasma parameters were chosen to be representative of the next step towards a fusion reactor (B = 5.3 T, ⟨ n_e⟩ = 1.0 × 10²⁰m^-3, R = 6.2 m). The modelling is performed with the TORAY code, launching a cone of electron cyclotron rays to account for a finite angular spectrum of injected waves. Current drive at each location is obtained with the Cohen linear model, which includes the effects of toroidal trapping, relativity and wave polarization. For the two launch locations, optimized central current drive efficiency is approximately equal. At plasma radius ρ = 0.835a for stabilization of the q = 2 NTM, the current density is 2.3 times greater and the integrated efficiency is 1.5 times greater for upper port launch relative to midplane launch. A broader range of frequencies for good current drive efficiency is obtained for upper port launch, and this will be reflected by a broader range of operating magnetic field at fixed electron cyclotron source frequency. Twenty megawatts of electron cyclotron power satisfy the criteria for control of NTMs.

A seed island is required for destabilizing the neoclassical tearing mode (NTM), which degrades confinement in long sustained, high confinement, high beta plasmas. The seed island formation due to an MHD event, such as a sawtooth crash, is investigated by applying an improved boundary layer theory of forced magnetic reconnection. This improved theory introduces non-constant-ψ matching and reveals the complicated nature of the reconnection described by two reconnected fluxes. In the initial evolution, these reconnected fluxes grow on a timescale consisting of the ideal timescale, the MHD event timescale and the resistive kink timescale. The surface current is negative, Δ'(t) < 0, to be consistent with NTM theory. The theory also yields a new integral equation which includes the typical timescale of the resistive kink mode and allows us to investigate the time evolution of the seed island at time t ∼ τ_A S^1/3.

A low aspect ratio quasi-axisymmetric stellarator, CHS-qa, has been designed. An optimization code has been used to design a magnetic field configuration with evaluations of the following physical quantities: quasi-axisymmetry, rotational transform, MHD stability and alpha particle collisionless confinement. It is shown that the electron neoclassical diffusion coefficient is similar to that of tokamaks for the low collisional regime. A self-consistent equilibrium with bootstrap current confirms the global mode stability up to 130 kA for an R = 1.5 m and B_t = 1.5 T device. The neoclassical plasma rotation viscosity is greatly suppressed compared with that of conventional stellarators. The engineering design was completed with 20 main modular coils and auxiliary coils, which provide flexibility of configuration in experiments for confinement improvement and MHD stability.

An integrated ingress of coolant event (ICE) test facility was constructed to demonstrate that the ITER safety design approach and design parameters for ICEs are adequate. The major objectives of the integrated ICE test facility are: to estimate the performance of an integrated pressure suppression system; to obtain validation data for safety analysis codes; and to clarify the effects of two phase pressure drops at the divertor and condensation in the suppression tank. The integrated ICE test facility simulates the current ITER components with a scaling factor of 1/1600. The TRAC-PF1 code is used to verify the integrated ICE experimental results. From the present study the effectiveness of the ITER pressure suppression system was verified experimentally and analytically, and then it was clarified quantitatively that the prediction accuracy of the TRAC-PF1 code is very high.

Bursts of turbulence associated with ELMs have been studied systematically in DIII-D with a multichannel phase contrast imaging (PCI) diagnostic, which is sensitive to the long poloidal wavelength components of the density fluctuations in the outer edge of the tokamak. A comparison of the temporal dynamics of the turbulence with the signature D_α signal from the divertor has revealed systematic differences between type I and type III ELMs: even though precursor fluctuations are sometimes seen before type I ELMs, the PCI signal level remains high until the peak in the D_α signal; by contrast, in type III ELMs the fluctuation burst precedes the D_α peak by 0.4-0.6 ms. Type I ELMs can generate `echoes', i.e. secondary bursts, in the scrape-off layer. Coherent modes are observed during type III ELMs only. The radial and temporal correlation structures and the spectral properties of the turbulence during the transient ELM phase have been reconstructed by averaging over multiple ELMs, in order to improve the statistical accuracy. ELM turbulence is found to share many properties with L mode turbulence, including the main qualitative features of radial wavenumber and frequency spectra and radial dispersion relations. However, features unique to ELM turbulence are also identified.

The paper highlights the JET work in physics and technology during the period of the JET Joint Undertaking (1978-1999), with special emphasis on what has been learned for extrapolation to a NEXT STEP device. - Global confinement scaling has been extended to high currents and heating powers. Dimensionless scaling experiments of ELMy H mode plasmas suggest that bulk plasma transport is gyro-Bohm and predict ignition for a device with ITER-FDR parameters. Experiments in which the plasma elongation and triangularity were varied independently show a strong increase of confinement time with elongation (τ_E ∼ κ_a^{0.8 ± 0.3}), thus supporting a basic design principle of ITER-FEAT. With the Pellet Enhanced Performance (PEP) mode, JET has discovered the beneficial effect of reversed magnetic shear on confinement, opening the possibility of advanced tokamak scenarios. - With a three stage programme of progressively more closed divertors, JET has demonstrated the benefits of divertor closure, in particular, of high divertor neutral pressure which facilitates helium removal. It has also shown that in detached (or semidetached) radiative divertor plasmas the average power load on the target plates of a NEXT STEP device should be tolerable but, in addition, that the transient power loads during ELMs could cause problems. - In 1991 JET has demonstrated the first ever controlled production of a megawatt of fusion power. More extensive D-T experiments in 1997 (DTE1) have established new records in fusion performance: 16 MW transient fusion power with Q_in = 0.62 (i.e. close to breakeven, Q_in = 1) and 4 MW steady state fusion power with Q_in = 0.18 for 4 s. DTE1 has also allowed a successful test of various reactor ICRF heating schemes and a clear demonstration of alpha particle heating, consistent with classical expectations. - JET has developed and tested some of the most important technologies for a NEXT STEP and a reactor, in particular the safe handling of tritium and the remote handling of large equipment as demonstrated by the Remote Divertor Exchange following DTE1.

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.