Technical Summary

Magnetic Confinement Theory and Modelling

Many of the papers were devoted to the investigation of transport processes, in particular to the toroidal momentum transport. Simulation gyro-kinetic codes have been improved in many countries and the number of available codes reaches now several tens. Numerical developments tend to follow the same trend as improvements on the computation power. The time scale for plasma simulations is now comparable with ion-ion collision time. It has been advocated that in order to improve predicting capability in characterising transport in ITER the near future advances are the combination of gyro-kinetic and fluid codes. Reports presenting works on stellarator configurations confirm that in these devices the neoclassical transport dominates but the influence of turbulent transport can play a role on improved confinement regimes and on the resilience of pressure profiles. Theoretical analysis and calculations showed that the Resonant Magnetic Perturbations (RMPs) can slowdown the plasma rotation increasing thus the probability of disruptions to occur.

The problems on Scrape off Layer (SOL) and divertor attracted a large number of theoretical works. As engineering solutions for divertor efficient operation are very challenging it calls for improved understanding of periphery plasma processes in order to provide new ideas for alternative schemes.

ITER and reactor studies have been presented and calculations confirm that ITER can achieve Q = 10 or larger. It has also been shown that the alpha-particle diffusion will be relatively small in ITER.

i) Plasma-material interactions, Divertors, Limiters, SOL; ii) Stability; iii) Wave-plasma interactions - current drive, heating, energetic particles; iv) Confinement

Most of the papers presented studies of great importance for ITER. Contribution to main stream research topics from small devices was also reported. Experiments on Edge Localised Mode (ELM) mitigation have increased the level of confidence for achieving robust ELM-controlled H-modes in ITER. ELM mitigation/suppression technique with edge ergodization is established by active control coils. Significant results of inter-machine experiments have been presented for ELM mitigation/suppression by impurity seeding is available (JT-60U), ELM pacing and mitigation by pellets has been demonstrated (JET), and divertor heat loads have been reduced by ELM control RMP (DIII-D).

Plasma facing components based on W have been experimented successfully in ASDEX Upgrade and JT-60U. In both devices, W accumulation was observed under certain conditions. Hydrogen and deuterium retention has been clear distinguished by using gas balance analysis during discharges and post mortem analysis to provide the reliable prediction of the tritium inventory in ITER (Tore Supra). Scaling laws for tritium retention in co-deposits have been developed based on systematic laboratory studies. The reduced D inventory has been confirmed in AUG with full-tungsten plasma facing components.

On the current drive and heating sub-topic, new results on the control of the sawtooth instability have been successfully demonstrated using real time control of the Electron Cyclotron (EC) launcher injection angle to modify the current profile around the q=1 surface. Disruption mitigation by massive gas injection and EC-Heating has been demonstrated (JET, JT-60U, C-Mod, FTU, AUG). Characteristics of thermal and current quenches have been investigated in tokamaks. During thermal quench on JET energy deposition timescale is identified to be 2-7 times longer than that of the plasma thermal energy collapse. Measurements in the early phase of the current quench in JT-60U suggest that the L/R model used for scaling law should be revised with a time dependent inductance.

Several papers presented ITER/DEMO oriented results. For steady state operation fully non-inductive current/current-less operation and ITER/DEMO relevant scenario has been demonstrated (Tore Supra, LHD, JT-60U, DIII-D, TCV, EAST, KSTAR, HT-7). Active control methods of Resistive Wall Mode (RWM), Neoclassical Tearing Mode (NTM) and other instabilities have been developed, based on understandings of their physical mechanisms (DIII-D, JT-60U, NSTX, LHD, RFX-mod, etc.).

Discharge with β > βN no-wall limit, has been confirmed by exceeding critical plasma rotation. Development of control scenarios have lead to higher βN regime (DIII-D, JT-60U, NSTX, RFX-mod). Thresholds of rotation (rotating shear), error field and ρ* to NTM have been found through multi-machine extrapolations.

Under the sub-topic of Plasma Wave Interaction & SOL/DIV physics improvements of plasma performance using lithium coating and suppression of impurity contamination in an ergodic layer have been reported (NSTX, FTU, TJ-II, T-10, LHD). Lithium coated PFCs (evaporation and Liquid Li limiter) reduce the recycling and impurity content, and modify the profiles of plasma parameters. As a result, improved confinement and extended periods of MHD quiescence are achieved whereas the resulting modification of the profiles seems to be different between devices. Impurity transport in the ergodic layer has been identified in LHD with broad stocastic divertor region. Using experimental data and 3-dimensional transport code, EMC3-EIRENE, impurity transport in the ergodic layer was investigated. Results show that the remnant islands in the stochastic layer have an impurity screening potential when the perpendicular energy transport dominates over the parallel one at high SOL densities.

The spatial profile of Alfvén eigenmode measured in DIII-D and the MHD calculations using NOVA code have shown good agreement. Alfvén eigenmode spectra and mode damping rate have been measured using the active excitation systems in Alcator C-Mod and MAST. It was demonstrated that the frequency analysis of fast ion loss detector signals in AUG enables the identification of the instability that induces fast-ion loss. Lost ions were also identified with trapped alpha particles in JET. A linear dependence of the loss intensity on Alfvén eigenmode amplitude has been directly derived from the experiments.

In helical plasmas, reversed shear Alfvén eigemodes and geodesic acoustic modes are excited simultaneously. The Alfvén eigemode that induced fast-ion loss was identified with the lost-ion orbit trace from the lost ion probe into the LHD plasma. A new kind of zonal flow, which is generated from the fast-ion orbit loss current induced by CHS fishbone mode was detected using the twin heavy ion beam probes.

Works on Electron Bernstein and higher harmonic fast waves have reported good progress. The power deposition profiles of the electron Bernstein wave heating measured in the WEGA stellarator were consistent with ray-tracing calculations.

Finally, on Confinement experiments the reports addressed four broad research areas:
i) Particle transport - peaking at low collisionality would improve fusion yield in ITER
ii) Plasma rotation - Is a critical issue for confinement and stability. New sources of torque, which can potentially increase rotation in ITER, are clearly identified, mode conversion current drive, lower hybrid, and non resonant magnetic perturbations. Also inward momentum pinch was identified on many experiments.
iii) Energy transport - Experiments addressing both projection and basic plasma physics of turbulence show among others that toroidal field ripple decreases the energy confinement.
iv) Pedestal and L-H transition - In addition to tokamaks, internal transport barriers were observed on stellarators and Reverse Field Pinches (RFPs). Clear evidence of critical temperature gradient was demonstrated. Progress in prediction of pedestal height is the key for accurate performance projections. Pedestal scaling with βP1/2 is more favourable for ITER.

Innovative Confinement Concepts and Operational Scenarios;

On Innovative Confinement Concepts, papers reported progress and excellent science in the areas of Field Reversed Configurations (simply-connected, no TF coils), Spheromak, Magnetic mirror (possible neutron source for component testing), Plasma focus (hot-dense plasma fusion neutron and X-ray sources), Magnetized Target Fusion for HEDLP (High Energy Density Laboratory Plasmas), Levitated dipole (high-beta stability and confinement). These include also reports on design study for direct energy conversion and alternative divertor materials and configurations.

Reports on Operational Scenarios sub topic showed that ITER scenario development has advanced significantly by means of improved start-up scenarios i.e. ELMing H-mode for Q=10 demonstrations and experimental advanced scenarios leading towards improved performance and long pulse operation. In this conference were reported results of Tokamak experiments reaching 100% bootstrap current. Also, studies on stellarator performance showed improvement by configuration optimization. Continued improvement in performance was achieved through optimization with respect to plasma shape, pressure profile, and current profile. Recently, work has started on the coupling the burning plasma core to boundary.

Fusion Technology

Reports were presented on new fusion research machines such as KSTAR - the First Nb3Sn based superconductor device with 94% availability reporting commissioning experience that might be valuable for ITER, and QUEST on developing steady state operation research by active wall temperature control. Reports were also presented on new machines with innovative features being designed or under construction such as IFMIF, W7X, KTM, FAST and NHTX to exploit steady state operation, materials, wall/plasma interface issues and new divertor designs, among others.

ITER Activities

Progress in many areas of basic plasma control and surface interaction physics has come from cross-machine collaboration. ITER discharge feasibility has been investigated by coordinated multi-machine effort.

Works on the physics of machine control for establishing a robust operating domain for the ITER 15 MA baseline scenario showed benefit from some changes on ITER PF coils positioning and current limits, and on divertor re-design. New designs for in-vessel coil in ITER for ELM control, Vertical Stability (VS) and RWM control were presented and discussed. Significant progress on ITER magnets was reported and progress on superconductor qualification is on-going. The diagnostics Working Group (WG) for ITER has made recommendations for some diagnostics to be omitted and for new diagnosis of some basic quantities to be included such as: high resolution neutron spectrometer, neutron calibration in divertor, infra-red thermography, collective thomson scattering, dust and tritium measurement to support licensing. Present R&D is on-going to improve front-end optics and develop especially in-situ optics cleaning techniques.

Strong progress was presented on heating and current drive (H&CD) sub-topic. The ITER launcher design as been chosen by H&CD WG. The operating parameters range for ITER were reached by the ECRH source development in Japan. Other works reported on ITER design of LHCD (EU) and on the ICRH concept tested on JET and Tore Supra.

The Neutral Beam Injection (NBI) source progress in Russian Federation foresees the construction and test of new intermediate step to ITER beam sources. Multi-Aperture Multi-Grid (MAMuG) Accelerator structure was chosen for ITER after CEA/JAEA joint test at JAEA Megavolt Test Facility. It shows better HV hold-off behavior than alternative SINGAP design and avoids the issue of large power (> 7MW) co-accelerated electrons in SINGAP. A few works addressed the pellet fuelling technology and proposals for disruption mitigation systems.

Papers on the development of the ITER divertor and first wall components addressed works on Tungsten characterisation, detritiation by RF-glow cleaning and on advanced divertor concepts for DEMO based on high temperature Tungsten, liquid and dust lithium. On the materials and blanket subject the Broader Approach (BA) activities have began addressing advanced neutron multipliers for higher stability at high temperature desired for pebble bed blankets in DEMO. Reports on advanced materials covered Reduced Activation Ferrite Martensic steels and SiCf /SiC composites. The US CTF concept was presented. A Wide variety of Reactor studies was presented such as low aspect ratio (A=2) slim central selenoid (JA), LHD-type Compact Stellarator (US), ST (Brazil) and Fission-Fusion (CNR) hybrids.

From the reports presented is possible to conclude that ITER is now benefiting from coordinated physics efforts and serious engineering activity solving many problems. However, fuelling still needs more studies.

Safety, Environmental and Economic Aspects of Fusion

Works on the economic aspects of fusion power plant showed that the most contributing parameter for an economically viable power plant is availability and thermodynamic efficiency. The development of materials is also important. On material qualification there is advantage in exceeding 5 MW.year/m2 neutron load as that could be a reasonable target for DEMO and no more than 15 MW.year/m2 for ultimate power plant.

Inertial Fusion Energy

Progress on Inertial Fusion Energy (IFE) research has been presented by Algeria, Australia, China, India, Italy, Korea, Poland, Spain, UK, U.S.A, Uzbekistan and Japan.

Recent progress on target was made on achieving ignition relevant areal densities above 200 mg/cm2, and on new compressing schemes that use multiple shock waves. New techniques were developed to time multiple hydrodynamic wave. Development of advanced ignition concetps were reported. New diagnostics were presented. The Multi-channel Multi-Imaging X-Ray Streak Camera was developed for observation of time-resolved two-dimensional X-ray images and time-resolved two dimensional temperature distributions.

The ignition phase, once demonstrated will become a mile stone for IFE and can be used as a firm bench mark for the forthcoming high gain and energy production developments. As for the central ignition approach the National Ignition Facility is almost completed and will provide X-ray based ignition with fusion gains of 10 in 2010. The Omega facility developed 60 beams with good uniformity and achieved 0.2 g/cm2 areal density with a D2 cryogenic target. As for the fast ignition approach sub-ignition experiments will take place soon at OMEGA EP and LFEX using state of art gratings to compress the laser pulse to around 10 ps. Status of development was reported on on-going projects to achieve beyond ignition parameters towards energy production: FIREX II (Japan), HiPER(EU), LIFE (USA) and Z pinch (USA).

The world wide, effort on inertial fusion energy appears coordinated and healthy. Many approaches are expected to be soon demonstrated or to be tested for ignition, high gain and energy production. Laser developments and target production are on track to be ready for the coming high repetition era.