Table of contents

Report on the IAEA Technical Committee Meeting held at Fukuoka, 25-29 October 1999.

A critical overview of the state of the art in fusion research is presented. The four main areas of activity are considered: physics, technology, engineering and safety. In physics, the trend is positive towards a better understanding of the suitable plasma regimes, which have to be confirmed through further experimentation on the operating machines. The quality of the engineering has already been demonstrated through the design and construction of a number of experimental machines, and a large database obtained from the design and operation of fission reactors is available. For safety very satisfactory results in analysing the impact of fusion on man and the environment are being achieved. In technology, an outstanding problem is that of finding low activation materials capable of resisting the harsh fusion environment for an adequate number of years. An intense 14 MEV neutron source is needed in order to allow the necessary developments.

The main programme of the Tore Supra tokamak is to investigate the route towards long pulse plasma discharges. Tore Supra is thus equipped with a superconducting toroidal magnet, a full set of actively cooled plasma facing components, and a heating and current drive capability based on high power RF systems connected to actively cooled antennas. After pioneering investigations using the LHCD system alone (2 min and zero loop voltage discharges), recent efforts have concentrated on finding scenarios to couple the two RF heating systems in order to perform high power, long duration discharges. To this aim, 6.5 MW, 25 s as well as 4 MW, 60 s discharges have been successfully achieved. At these high power levels, the plasma-wall interaction becomes a critical issue, and recycling fluxes must be controlled to maintain density and to avoid plasma contamination. All these results contributed to the validation of the upgrade of the Tore Supra first wall components (CIEL project) scheduled for 2000.

Significant progress has been made on the HT-7 superconducting tokamak recently. Repeatable plasma discharges for up to 10.7 s have been routinely obtained, which nearly reached the present limit of the PF system. LHCD experiments have been successfully carried out. A full LH wave current drive for up to 3 s has been achieved. A quasi-steady-state H mode was observed with a very high plasma density for LHCD. ICRF waves were used for wall conditioning, startup, and fast wave and ion Bernstein wave heating. A new RF conditioning technique is shown to be a very powerful means of conditioning future large superconducting tokamaks. The physical design of HT-7U has been nearly finished. The engineering design and some R&D have started. A 2:3 model of the TF coil and a 1:4 model of the central solenoid with a cable in conduit conductor cable have been finished. Testing of the coils will start soon. The engineering design is under way in accordance with the schedule.

SST-1 is a steady state superconducting tokamak used to study the physics of plasma processes in tokamaks under steady state conditions and to gain knowledge about technologies related to steady state tokamak operation. Major subsystems of SST-1 are described and the present status of the SST-1 project is presented.

The stellarator Wendelstein 7-X (W7-X) is presently under construction at Greifswald, Germany, and the start of operation is planned in 2006. W7-X is a large `advanced stellarator' of the HELIAS type (R = 5.5 m, a = 0.55 m, B₀ = 3 T, five periods, moderate shear and variable rotational transform 5/6 ⩽ ι ⩽ 5/4 at the boundary) with the aims of demonstrating the reactor potential of this stellarator line in steady state operation close to fusion relevant parameters. The capability of stationary operation requires the realization of a superconducting magnet system consisting of 50 modular coils and 20 planar coils, the operation of a 140 GHz ECR CW heat source of 10 MW, the installation of a divertor to handle the power and particle flux, and to limit the impurity fraction to tolerable levels. Additional heating schemes, ICRF and NBI, will be provided for flexible experimentation.

Over the last thirty years, there have been increasingly ambitious theories and experiments along the theme of exercising increased control over plasma by means of RF waves. The diversion of energy from energetic alpha particles to waves (alpha channelling) is such an attempt at detailed control over plasma behavior. The effect could accelerate progress towards an economical deuterium-tritium tokamak reactor. Recent simulations and experiments on TFTR support certain separate building blocks that taken together might produce the desired effect.

The fundamental differences between inductive and non-inductive current buildup are clarified and the associated timescales and other implications are discussed. A simulation is presented whereby the plasma current in a low aspect ratio torus is increased primarily by self-generated bootstrap current, with only 10% coming from external current drive. The maximum obtainable plasma current from this process is shown to scale with toroidal field strength. The basic physics setting the timescales can be obtained from a 1-D analysis. Comparisons are made between the timescales found here and those reported in the experimental literature.

High performance discharges are routinely obtained on JET with low or reversed magnetic shear (s = (r/q)dq/dr), and the potential for steady state operation of such discharges is under investigation. With the use of the proper heating and fuelling, these `optimized shear' (OS) discharges exhibit an internal transport barrier (ITB), resulting in a strong peaking of the pressure profile, and thus in high fusion performance. These regimes have been extensively studied during the last (DD and DT) JET campaigns in order to promote this type of scenario as the basis for `advanced tokamak' operation. A review is given of the highest performance achieved on JET OS discharges during the last experimental campaigns, in both DD (up to 5.6 × 10¹⁶neutrons/s) and DT operation (fusion power up to 8.2 MW, n_i0T_i0τ_E up to 10²¹ m^-3 keV s). The role of the plasma edge is pointed out, as the power required to trigger an ITB is often higher than the H mode power threshold, leading to double barrier regimes. The presence of an H mode pedestal both modifies the ITB and induces edge bootstrap and ELM activity, which should be controlled to prolong such discharges. The operational procedure of optimization is then discussed, addressing the problems of ITB formation (power threshold, timing of the main heating phase, i.e. optimization of the target q profile, influence of the heating scheme, electron versus ion ITBs), ITB evolution (expansion of the ITB footpoint, H mode formation) and ITB termination (disruptive and/or `soft' MHD limits). Finally, the crucial problem of the route to steady state for such OS discharges is addressed, both in terms of ITB sustainment and control within the stability domain and in terms of edge pedestal control by means of impurity injection. The impurity behaviour is found, and examples of high performance discharges sustained for several energy confinement times are given (β_N = 1.95, H₈₉ = 2.3, P_fusion^eq∼10 MW, Q_DT^eq∼0.4 sustained for ∼3 s). Extrapolation towards fully non-inductive current drive is discussed.

Recent experimental results for steady state high performance plasma on JT-60U are described, emphasizing plasma control, especially active control of profiles. In reversed shear (RS) plasma, a DT equivalent fusion gain of 0.5 was sustained for 0.8 s (∼ the energy confinement time) by controlling plasma β with a newly developed stored energy feedback control technique. An internal transport barrier (ITB) with an RS configuration was successfully maintained by means of LHCD in full current drive quasi-steady state for 4.7 s. The viability of a steady state RS configuration sustained by a large bootstrap current fraction has been demonstrated in ELMy H mode edge RS plasmas for the first time. In high-β_p ELMy H mode plasmas, an integrated core plasma performance with H factor of 2.56, β_N = 2.4, and bootstrap current fraction and non-inductively driven current fraction close to the ITER requirements was achieved at I_p = 1.5 MA in a near full current drive state by negative ion based NB (N-NB) injection. A rudimentary simulation of a reactor operation with simultaneous feedback control of both the core and divertor plasmas has been demonstrated in ELMy H mode plasma. Aiming at active control of the pressure gradient at the ITB in RS plasmas, a feedback control system of the electron temperature gradient has been developed and optimization of the discharge scenario using this system has recently started.

Recent progress in the development of long pulse, high performance discharges on the DIII-D tokamak is reviewed. This is highlighted by a discharge achieving simultaneously a β_N H of 9, bootstrap current fraction of 0.5 and non-inductive current fraction of 0.75, sustained for 16 energy confinement times. The physics challenge has changed in the long pulse regime. Non-ideal MHD modes limit the stability, fast ion driven modes may play a role in fast ion transport, which limits the stored energy, and plasma edge behaviour can affect the global performance. New control tools are being developed to address these issues.

A review is presented of the scenarios and physics of advanced tokamak discharges and the associated technical enhancements leading towards enhanced performance and steady state operation in ASDEX Upgrade. A stationary advanced tokamak scenario with internal transport barriers (ITBs) in combination with an H mode barrier and weak shear (q_min ≈ 1) was maintained for 40 confinement times and several internal skin times with H_ITERL-89Pβ_N ≈ 5. By raising the triangularity of the plasma shape the performance was increased up to H_ITERL-89Pβ_N ≈ 7.2, but β is still limited by neoclassical tearing modes. The density was raised to close to 50% of the Greenwald density either by edge gas fuelling, causing an increase of the threshold power to sustain an ITB and a decrease of Z_eff below 2, or by improved core particle confinement with more triangular plasma shapes without changing the ITB onset conditions. Sufficient helium pumping and no temporal impurity accumulation was observed despite peaked impurity density profiles. MHD modes contribute to making the shear profile stationary. In the ITB/H mode scenario (1,1) fishbones clamp the q value to the vicinity of 1 and avoid sawteeth, while in ITB scenarios with reversed shear (q_min ≈ 2) (2,1) fishbones can clamp the current profile development near the q = 2 surface without causing energy confinement to deteriorate. Double tearing modes act similarly, but lead to substantial confinement losses. Applying central ECRF heating and current drive to beam heated reversed shear ITB discharges shows a substantial effect on MHD stability, affecting the passage of the q profile through q_min = 2 and degrading or prolonging the reversed shear phase, depending on the current drive direction. Moreover, reactor relevant T_e ≈ T_i operation with temperatures in excess of 10 keV was achieved with internal transport barriers for both electrons and ions simultaneously. For current profile control ECCD will be supplemented by on-axis fast wave ICCD and off-axis current drive up to 400 kA using NBI (available in 2001) and ICRF mode conversion. Stationary discharges with reversed shear and q_min > 2 should then be possible at a plasma current of 1 MA according to power deposition, current drive and transport calculations.

The Large Helical Device is the world's largest heliotron type helical system, with the plasma confining magnetic field being generated by only external superconducting coils. One of the main objectives of the LHD project is to sustain high temperature plasmas for a long time in steady state. The plasma vacuum vessel and the divertor are water cooled, and a heat load of 3 MW can be removed continuously. The NBI, ECH and ICRF heating systems, diagnostic instruments and data acquisition system are designed for long pulse operation. The present status of these systems and the recent experimental results of long pulse operation are reviewed. A steady state discharge with NBI was obtained for 35 s. The ECH discharge duration was extended to 120 s with a duty factor of 95%. Plasma sustainment by ICRF alone was achieved for 2 s. The performance of these long pulse operations is summarized.

Plasma control concepts are introduced which are likely to change when going from a 500-1000 s tokamak pulse to the discharges in a steady state tokamak, whose more economical design relies on the reduced number of thermal and mechanical stress duty cycles.

A steady state plasma with high performance and high current drive efficiency is reported. In 2.45 GHz LHCD plasmas T_i is studied as a function of n_e at the edge of the high ion temperature (HIT) window. Different characteristic timescales are found for T_i and n_e to enter the HIT regime and the observed hysteresis behaviour of T_i with respect to n_e is attributed to this difference. The electromagnetic emission (<3.5 GHz) is studied in order to understand ion heating mechanisms in the HIT regime. The spectrum shows several sidebands whose peak frequencies correspond to the ion plasma frequency. The spectral narrowing of the width of the sideband shows a clear correlation with ion heating. In 8.2 GHz LHCD plasmas an enhanced current drive (ECD) regime where both current drive efficiency η_CD( = _eI_CDR₀/P_LH ∼1 × 10¹⁹ A m^-2/W) and energy confinement time τ_E (∼8-10 ms) are simultaneously improved is obtained at an _e of 4.3 × 10¹³ cm^-3 and B = 7 T under full current drive conditions. There exists a certain threshold power above which the ECD transition occurs. A hysteresis of η_CD is found around the threshold power, which is explained by the different characteristic time for the ECD transition in power rampup and rampdown schemes. Current profile control experiments are performed by using two opposite travelling LHWs. Current compensation (ΔI_CD/I_CD < -10%) is clearly seen when the backward (BW) travelling LHW (8.2 GHz) is added to a target plasma whose current is driven by a forward travelling LHW (8.2 GHz). As the BW wave power is increased, however, the current tends to flow in the forward direction. The mechanisms of this non-linear behaviour of the driven current with respect to the BW wave power are discussed.

After a decision by the ITER parties to investigate the possibility of designing a reduced cost version of ITER several possible machine layouts with different aspect ratios were studied. Relatively early in this process it became clear that there is no significant cost difference between different aspect ratios and that there is a maximum realistically possible aspect ratio for a machine with 6 m major radius and rather high plasma shaping. Following this study a machine with an intermediate aspect ratio (3.1) called the ITER Fusion Energy Advanced Tokamak (ITER FEAT) was chosen as the basis for the outline design of a reduced cost ITER. Several potential steady state scenarios can be investigated in ITER FEAT, i.e. monotonic or reversed shear at full or reduced minor radius. In addition, so-called hybrid discharges, are feasible where a mixture of inductive and non-inductive current drive as well as bootstrap current allows long pulse discharges of the order of 2500 s. The β_N values and H factors required for these discharges are in the same range as those observed on present machines, which provides confidence that such discharges can be studied in ITER FEAT. However, due to uncertainties in physics knowledge, for example the current drive efficiency off-axis, it is impossible at present to generate a completely self-consistent scenario taking all boundary conditions, for example engineering or heating system constraints, into account. In addition, all of these regimes have a potential problem with divertor operation compatibility (low edge density) and with helium exhaust which has to be addressed in existing experiments. For the engineering design of the in-vessel components and for the balance of the plant there is practically no difference between inductive (500 s) and steady state operation. However, the choice of heating systems and the distribution of power between them will be strongly influenced by the envisaged steady state scenarios.

For a fusion power plant to be economically and technologically attractive, it should be as compact as possible and capable of `steady state' operation. One approach is based upon the spherical tokamak (ST) concept. This configuration features many of the qualities of the conventional aspect ratio tokamak, such as good confinement and MHD stability, together with a number of highly promising features for the realization of a cost effective, steady state fusion power core. In particular, it allows high β capability due to the high I/B achievable at low A and strong, natural shaping of the configuration, together with the possibility of approaching 100% pressure driven currents through high natural elongation and access to second stability to ideal ballooning modes. The physics insights gained from experiments on START are discussed and the theoretical modelling work summarized, describing how the beneficial properties of STs can be combined in steady state ST power plant designs such as the Culham STPP concept. Where appropriate, we shall refer to the two new mega-amp machines MAST and NSTX, both designed for a plasma duration approaching the plasma current relaxation time and also other new, smaller, STs, designed to expand the knowledge gained so far.

The basics of using the cable in conduit conductor (CICC) concept in new projects for magnetic fusion devices as well as the need to improve the CICC design are discussed. Preference is given to a regular inner structure with respect to the usually considered loose one. Optimal solutions for the cooling scheme, the way to avoid all coil pre-assembly tests, the support structure, vacuum vessel (VV) and cryostat design that could provide the best access to the device, load transfer from VV to cryostat and other design features of new projects are discussed.

A combined model for current profile control and MHD stability analysis has been used to identify stable operating modes near the ideal stability limit (β_N ≃ 3) in the Alcator C-Mod tokamak. These discharges are characterized by relatively high fractions of bootstrap current (f_BS = 0.70) and non-monotonic profiles of the safety factor with q_min > 2. In the absence of a conducting shell, stability was determined by the onset of the low (n = 1) external kink mode. In these studies, current profile control in the plasma periphery (r/a ≳ 0.5) was provided by 2.5-3.0 MW of LHCD power. Internal and edge transport barriers were introduced into the model calculations in the form of density transitions. Excellent wave accessibility and absorption were still found in the presence of an H-mode-like edge density barrier. However, the presence of these barriers resulted in about a 10% decrease in the stability limit, from β_N ≃ 3 to β_N ≃ 2.7.

The status of modelling work focused on developing the advanced tokamak (AT) scenarios in DIII-D is discussed. The objective of the work is twofold: (a) to develop AT scenarios with ECCD using time dependent transport simulations, coupled with heating and current drive models, consistent with MHD equilibrium and stability; and (b) to use time dependent simulations to help plan experiments and to understand the key physics involved. Time dependent simulations based on transport coefficients derived from experimentally achieved target discharges are used to perform AT scenario modelling. The modelling indicates that off-axis ECCD with approximately 3 MW absorbed power can maintain high performance discharges with q_min > 1 for 5-10 s. The resultant equilibria are calculated to be stable to n = 1 pressure driven modes. The plasma is well into the second stability regime for high-n ballooning modes over a large part of the plasma volume. The role of continuous localized ECCD is studied for stabilizing m/n = 2/1 tearing modes. Progress towards validating current drive and transport models, consistent with experimental results, and developing self-consistent, integrated high performance AT scenarios is discussed.