Table of contents

In the general area of confinement improvement and concept advancement, recent results in the Madison Symmetric Torus (MST) reversed-field pinch (RFP) include good confinement of both thermal and large-orbit ions and near doubling of total beta to 26% with deuterium pellet injection. Current profile control enables MST to reduce stochastic transport and achieve tokamak-like confinement. In standard RFP operation, substantial MHD tearing mode activity results in stochastic transport and an energy confinement time of about 1 ms in MST. Application of inductive current profile control reduces MHD activity and accompanying stochasticity, improving confinement by about a factor of ten. Previous work concentrated on electron confinement in improved-confinement RFP operation. Recent work confirms that ions are also well confined, and that high beta and improved confinement can be achieved simultaneously.

Collapse events at the q = 2 surface occurring during the mode locking process of an m/n = 2/1 tearing mode have been observed on TEXTOR. The plasma confinement within the q = 2 surface collapses without much influence on the width of the island (O-point). With an external rotating resonant magnetic perturbation (RMP) field induced by the dynamic ergodic divertor, secondary structures moving near the separatrix of the primary large 2/1 island with the same frequency as the RMP have been observed after the collapse events. The plasma confinement recovers when the secondary structures vanish.

Observations on Alcator C-Mod confirm and extend previous measurements from ASDEX Upgrade (AUG) and JET, where density peaking in H-mode was seen to increase as collisionality was lowered. In the earlier work, peaking was also strongly correlated with n_e/n_G, the ratio of the plasma density to the density limit, and it was not possible to eliminate the latter parameter as the principal scaling variable. This led to divergent predictions for the density profile on ITER. The addition of C-Mod data allows the correlation to be broken, strengthening the case for collisionality as the controlling parameter. At the lowest collisionalities, C-Mod attained peaking factors n_e(0)/⟨n_e⟩ approaching 1.5 with T_i = T_e, high edge neutral opacity and no core fuelling. Taken together, data from the three devices strongly suggest that density peaking factor on ITER in ELMy H-modes will be in the range 1.4–1.6.

A quasi-steady-state alternating current operation assisted by lower hybrid wave (LHW) was achieved on a HT-7 superconducting tokamak with plasma current of I_p = 125 kA, line-averaged density of 1.5 × 10¹⁹ m⁻³, electron temperature of T_e = 500 eV and 30–50 s plasma duration. Plasma current was sustained and smoothly transferred from one direction to the other without loss of ionization. Plasma position control, LHW assistance, strong gas puffing and good wall condition are the key issues to have a smooth transition of plasma current. Our modelling results show that current reversal equilibrium configuration with two oppositely flowing currents in the high-field-side and the low-field-side during current reversal exists. This is in agreement with experimental measurements.

An increase in the ion temperature due to transport improvement has been observed in plasmas heated with high-energy negative-ion-based neutral beam injection (NBI), in which the electrons are dominantly heated, in the Large Helical Device. When centrally focused electron-cyclotron-resonance heating is superposed on the NBI plasma, the ion temperature is observed to rise, accompanied by the formation of the electron internal transport barrier (electron-ITB). In the electron-ITB plasmas, an increase in a positive radial electric field is observed, and the transport analysis indicates that the ion transport in the half-radius region is improved with a reduction of the anomalous transport. Thus, this ion temperature rise is ascribed to the ion transport improvement by the transition to the neoclassical electron root. In high-Z plasmas, the ion temperature is increased with an increase in the ion heating power and reaches 13.5 keV. The central ion temperature increases with an increase in a gradient of the electron temperature in an outer plasma region of ρ = 0.8, suggesting the ion transport improvement in the outer plasma region induced by the neoclassical electron root. These results indicate the effectiveness of the electron-root scenario for obtaining high-ion temperature plasmas in helical systems.

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.

In the modified RFX experiment (RFX-mod) external magnetic field coils and a close fitting thin conductive shell control radial magnetic fields. In the so-called virtual shell (VS) operation, radial field zeroing at the thin shell radius is stationary provided by the feedback-controlled coils. First experiments on RFX-mod proved the capability of the active scheme to steadily reduce the radial magnetic field. Furthermore it has been found that such edge magnetic field control extends its beneficial effects to the whole plasma. With respect to the old RFX, where magnetohydrodynamic modes amplitude was controlled by the use of a passive thick conductive shell, a stationary 2- to 3-fold reduction of the B_r field amplitude in the core is obtained. The reduction of field fluctuations positively reflects on confinement. In fact, a strong reduction of the loop voltage is observed and correspondingly a 3-fold increase in pulse length is achieved by using the same poloidal flux swing. Temperature and particle measurements confirm the improved confinement properties of the VS operation. With a lower ohmic input power, higher electron temperature and lower particle influx are measured. Particle and heat transport have been studied by means of a 1D code. Local power balance was used to compute the heat conductivity profile: for the VS discharges a lower conductivity over a significant region of the plasma is found. The improved properties of RFX-mod VS operation provide a better confinement scaling in terms of plasma current. The results show that compared with the thick shell configuration, a significant confinement improvement can be obtained under stationary conditions by actively controlling the plasma magnetic boundary.

For the modelling of hot reactor-grade plasmas a way to compute local electron cyclotron wave losses efficiently and with good accuracy is required. Thereby the approach to calculating the electron cyclotron absorption coefficient is an essential element. A comparison of three available analytical forms (a 'quasi-exact' one and the asymptotic forms due to Trubnikov and to Robinson) shows that, overall, the use of Robinson's formula is the most suitable choice.

An integrated view of disruption dynamics on internal plasma and associated electromagnetic structures with a high time resolution is presented in a systematic way. The main results are as follows: (i) observation of electron temperature oscillation due to rotation of magnetic islands of m(poloidal)/n(toroidal) = 3/2 mode numbers before the current quench, (ii) finding of rapid pump-out (∼20 µs) of plasma current and particles from the central core region to the edge just at the current quench start time and then returning to a peaked profile during the following slow decay phase and (iii) a new proposal for the evaluation of current quench time and the electron temperature dependence of the current quench time, the so-called τ/S scaling.

The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350 g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, namely, ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is ∼10–20%. This is larger than the ∼3–4% deduced from post-mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes are developed.

Recent high-β DIII-D (Luxon J.L. 2002 Nucl. Fusion42 64) experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable when the plasma rotation is lowered to a fraction of a per cent of the Alfvén frequency by reducing the injection of angular momentum in discharges with minimized magnetic field errors. Previous DIII-D experiments yielded a high plasma rotation threshold (of order a few per cent of the Alfvén frequency) for RWM stabilization when resonant magnetic braking was applied to lower the plasma rotation. We propose that the previously observed rotation threshold can be explained as the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Resonant braking can also occur naturally in a plasma subject to magnetic instabilities with a zero frequency component, such as edge localized modes. In DIII-D, robust RWM stabilization can be achieved using simultaneous feedback control of the two sets of non-axisymmetric coils. Slow feedback control of the external coils is used for dynamic error field correction; fast feedback control of the internal non-axisymmetric coils provides RWM stabilization during transient periods of low rotation. This method of active control of the n = 1 RWM has opened access to new regimes of high performance in DIII-D. Very high plasma pressure combined with elevated q_min for high bootstrap current fraction, and internal transport barriers for high energy confinement, are sustained for almost 2 s, or 10 energy confinement times, suggesting a possible path to high fusion performance, steady-state tokamak scenarios.

Alfvén eigenmodes (AEs) excited by thermonuclear α-particles can degrade the heating efficiency by spatial redistribution of the resonant α-particles. Changes of the orbit invariants in phase space by collisions and interactions with other waves, such as magnetosonic waves during ion cyclotron resonance heating (ICRH), lead to changes in the phase between the α-particles and AEs, causing a decorrelation of the interactions and stronger redistribution of the α-particles. Cyclotron interactions increase the decorrelation of the AE interactions with the high-energy ions and hence a stronger radial redistribution of the high-energy α-particles by the AEs. Renewal of the distribution function by thermonuclear reactions and losses of α-particles to the wall lead to a continuous drive of the AEs and a radial redistribution of the α-particles. The condition for excitation of AEs is shown to depend on the heating scenario where heating at the low field side creates a significant population of high-energy non-standard orbits which drive the modes. The redistribution results in a reduction in the averaged α-particle energy and a degradation of the heating efficiency. The effect on the distribution function in the presence of several unstable modes is not additive and the particle redistribution is found to saturate with an increasing number of modes.

This paper reports the progress of R&D towards the ITER neutral beam (NB) system at the Japan Atomic Energy Agency (JAEA) during 2004–2006. In the accelerator development, an H⁻ ion beam of 146 A m⁻² was accelerated up to 836 keV in a vacuum insulated accelerator with five acceleration stages. This is the first production of a high power density beam at MeV range energy, which is more than twice as high as those of the existing NB systems for large experimental reactors such as JT-60U and LHD. The ITER NB system requires a high voltage (HV) bushing as an electric feedthrough to the vacuum insulated accelerator from the gas insulated power supply. A full-size (1.56 m in diameter and 0.292 m in height) insulator ring made of alumina ceramic has been successfully fabricated for HV bushing, as a result of the technological development of a new cold iso-static press.

Long pulse operations of JT-60U negative ion sources with a large extraction area of 0.45 × 1.1 m² are in progress. From the two ion sources, D⁻ ion beams of 21 A (90 A m⁻²), 320 keV have been produced for 21 s. As a consequence, injection of 3.2 MW D⁰ beams was achieved. Thus, the long pulse operation of the negative ion based NB system has been demonstrated with multi-tens of ampere of D⁻ ions extracted from the ion extraction area relevant to the ITER NB design (0.6 × 1.5 m²). For long pulse operation of such powerful beams, the negative ion uniformity over the wide extraction area is an essential issue to avoid excess power loading on the accelerator grids. By adopting a new magnetic configuration with the so-called 'tent-shaped filter', the uniformity of the JAEA 10 A source (extraction area: 0.13 × 0.22 m²) was improved by 8%, which fulfilled the ITER requirement.

Various gain models have shown the potentially great advantages of fast ignition (FI) inertial confinement fusion (ICF) over its conventional hot spot ignition counterpart (e.g. Atzeni S. 1999 Phys. Plasmas6 3316; Tabak M. et al 2006 Fusion Sci. Technol.49 254). These gain models, however, all assume nearly uniform density fuel assemblies. In contrast, conventional ICF implosions yield hollowed fuel assemblies with a high-density shell of fuel surrounding a low-density, high-pressure hot spot. Hence, to realize fully the advantages of FI, an alternative implosion design must be found which yields nearly isochoric fuel assemblies without substantial hot spots. Here, it is shown that a self-similar spherical implosion of the type originally studied by Guderley (1942 Luftfahrtforschung19 302) may be employed to yield precisely such quasi-isochoric imploded states. The difficulty remains, however, of accessing these self-similarly imploding configurations from initial conditions representing an actual ICF target, namely a uniform, solid-density shell at rest. Furthermore, these specialized implosions must be realized for practicable drive parameters and at the scales and energies of interest in ICF. A direct-drive implosion scheme is presented which meets all of these requirements and reaches a nearly isochoric assembled density of 300 g cm⁻³ and areal density of 2.4 g cm⁻² using 485 kJ of laser energy.

In order to achieve the highest plasma pressure limits in ITER, resistive wall kink mode stabilization is required. A novel resistive wall mode linear observer and feedback controller designed using model reduction and optimal control theory and employing only proportional gain are described here that allow operation of ITER up to C_β = 86% of the ideal wall limit using the present design external control coils. The full VALEN finite element ITER model containing ∼3000 modes was reduced to a minimum of 8 modes making real-time controller implementation possible. We find an order of magnitude reduction of the required control coil current and voltage in the presence of white noise from the no-wall limit to the optimal feedback system performance limit as compared with a traditional, classical controller.

Pellets injected into type-I ELMy H-mode discharges are known to trigger edge-localized modes (ELMs). In order to understand the underlying processes the triggering mechanism was investigated in this paper. The major questions of the investigations to be answered were: at which magnetic surface was the ELM initiated and what was the corresponding perturbation caused by the ablating pellet? During the investigations the natural ELM cycle was probed by injecting pellets from the high field side of the ASDEX Upgrade tokamak with significantly lower frequency than the natural ELM frequency. To determine the location of the seed perturbation of the ablating pellet triggering an ELM, the dynamics of the triggered ELMs was linked to the time history of the pellet position in the plasma. The ELM onset was determined by analysing magnetic pick-up coil signals and its delay relative to the time when the pellet crossed the separatrix was measured as a function of the pellet velocity. Supposing that to trigger an ELM a pellet has to reach a certain magnetic surface of the plasma independently of its mass and velocity, the most probable location of the seed perturbation was found to be at the middle of the pedestal—at the high plasma pressure gradient region. The onset of the MHD signature of the ELMs was detected about 50 µs after the pellet reached the seed position. According to our observations ELMs can be triggered either by the cooling of the pedestal region causing sudden increase of the pedestal plasma pressure gradient driving the plasma to the unstable region of the ballooning mode or by the strong MHD perturbation triggering an instability developing into an ELM.

The energy leakage probability of charged particles from an ICF pellet has been estimated earlier using a small angle binary collision approximation for Coulomb scattering from electrons and ions. While this is the most predominant energy loss mechanism, elastic nuclear scattering is important for high density pellets at higher temperatures. In this paper, we generalize the calculation of energy leakage probability to include nuclear scattering, large angle Coulomb scattering and collective plasma effects. In general, these effects reduce the thermalization distance in the plasma and increase the fraction of energy deposited to ions. We also develop a simple approach for energy deposition by neutrons due to nuclear interaction with the ions. The same model is then used to re-evaluate the concept of internal tritium breeding in high density ICF pellets. It is found that tritium breeding improves significantly in comparison with earlier estimates.

In the advanced reversed-field pinch (RFP), current profile control (CPC) enables energy confinement time and poloidal beta to increase substantially as compared with the conventional RFP due to reduced magnetic field stochasticity. Numerical simulations using the three-dimensional non-linear resistive MHD-code DEBSP are performed showing that the poloidal beta is not limited to the m = 0 stability criterion β_θ < 1/2. Instead, as tearing modes are diminished, it may approach unity. The beta criterion is theoretically analysed and a new, more general, criterion is derived. Analytic estimates of the resistive tearing and g-mode growth rates are derived for m = 0, and it is shown that both tearing and g-mode growth rates decrease significantly as CPC is employed. Furthermore, quasi-steady state operation with increased confinement due to active control of the current profile is numerically demonstrated for the advanced RFP for a scenario with β_θ < 1/2.

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ∼10–20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

This paper presents the results obtained with a set of gyrokinetic codes based on a semi-Lagrangian scheme. Several physics issues are addressed, namely, the comparison between fluid and kinetic descriptions, the intermittent behaviour of flux driven turbulence and the role of large scale flows in toroidal ITG turbulence. The question of the initialization of full-F simulations is also discussed.

The improvement of core electron heat confinement has been realized in a wide range of helical devices such as CHS, LHD, TJ-II and W7-AS. Strongly peaked electron temperature profiles and large positive radial electric field, E_r, in the core region are common features for this improved confinement. Such observations are consistent with a transition to the 'electron-root' solution of the ambipolarity condition for E_r in the context of neoclassical transport, which is unique to non-axisymmetric configurations. Based on this background, this improved confinement has been collectively dubbed 'core electron-root confinement' (CERC). The thresholds for CERC establishment are found for the collisionality and electron cyclotron heating power. The magnetic configuration properties (e.g. effective ripple and magnetic islands/rational surfaces) play important roles for CERC establishment.

Internal transport barriers (ITBs) can be routinely produced in enhanced D_α (EDA) H-mode discharges on the Alcator C-Mod tokamak by putting the minority ion cyclotron resonance layer at |r/a| ≥ 0.5 during the current flat top phase of the discharge. These ITBs are characterized by density peaking at constant temperature and are therefore both particle and energy transport barriers. The ITB formation appears to result from widening the region near the magnetic axis in which toroidal drift modes are stable, allowing the Ware pinch to peak the density profile. Experimental evidence shows that shifting the ICRF resonance off-axis results in a local flattening of ion and electron temperature profiles. TRANSP calculations of ion temperature profiles support this experimentally observed trend. Stability analysis of ion temperature gradient (ITG) and electron temperature gradient modes at times before ITB formation is done using the linear gyrokinetic code GS2. These gyrokinetic calculations find that the most unstable modes in the C-Mod EDA H-mode core, prior to ITB onset, are the toroidal ITG driven type. These modes are suppressed in the ITB region through a temperature gradient reduction when the ICRF resonance is shifted off-axis.

This paper addresses a basic equilibrium/confinement question whether a field reversed configuration (FRC) with a balanced amount of magnetic and thermal energies is sustained solely by inductive magnetic flux (magnetic energy) injection of a centre solenoid coil. A new finding is that preferential injection of magnetic energy causes a significant increase in thermal energy of the FRC to maintain high-beta equilibrium. The FRC plasma is concluded to have the robustness needed to self-adjust the plasma heating power depending on the magnetic energy injection. In the flux ramped-up phase, i.e. in the phase where the applied inductive electric field did not penetrate into the core of the plasma, the J × B compression (pinch effect) was considered to be dominant for heating the FRC. On the other hand, after the applied inductive electric field penetrated into the core region the resistivity of the FRC increased up to 10–20 times larger than the classical spitzer resistivity. This anomalous heating was the most probable cause for sustainment of the high-beta FRC in this phase. A quasi-steady state of the high-beta (volume averaged beta ∼0.6–0.7) FRC for 0.15 ms (∼energy confinement time of conventional FRCs) was achieved by the use of the present low power centre solenoid current drive. This successful result leads us to the next stage steady sustainment experiment of an FRC by increased capacitor bank energy.

A generalized Newcomb equation that incorporates inertial contributions from equilibrium shear flows as well as finite β contributions is derived in a cylindrical geometry. It is shown from numerical solutions of this equation that the stability parameter Δ' can be significantly influenced by flow as it becomes a sensitive function of the global profiles of the magnetic field and flow velocity. These results can have important implications for the nonlinear evolution of magnetic islands in the presence of flows.

Based on the earlier work for a single magnetic island and a weak stochastic magnetic field (Yu 2006 Phys. Plasmas13 062310), the heat diffusion across a local fully stochastic magnetic field is studied numerically and compared with the analytical theories. The ratio between the parallel and the perpendicular heat diffusivity, χ_||/χ_⊥, is found to be an important parameter in characterizing the transport. With the increase in χ_||/χ_⊥, the transport is dominated first by the additive effect of individual islands and then by the field ergodicity and χ_||.

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 10¹⁹ m⁻³ and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 10¹⁹ m⁻³ and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

Collisionless damping of the geodesic acoustic mode (GAM) in helical magnetic configuration is investigated by a drift-kinetic simulation. We utilize the neoclassical transport code 'FORTEC-3D', which solves the drift-kinetic equation based on the δf method, to study how the magnetic configuration affects the collisionless damping of GAMs. In a recent analytical study, in which a simplified expression for helical magnetic field was used, it was found that the helical ripples strengthen the damping rate of the GAM oscillation. In this paper, the change in the damping rate is investigated in detail for realistic magnetic field configurations of the Large Helical Device. It is found that the GAM damping is faster and the frequency is lower as the magnetic axis is shifted inwards. This result suggests the possibility of controlling both the neoclassical transport level and the GAM oscillation, or the zonal flow, in helical plasmas. The collisional effect on GAM damping is also investigated, however, it is found that the GAM oscillation in helical plasmas is insensitive to the collisional effect because the enhancement of collisionless damping by the magnetic field ripples is very strong.

The International Stellarator Confinement Database (ISCDB) is a joint effort of the helical device community. It is publicly available at http://www.ipp.mpg.de/ISS and http://iscdb.nifs.ac.jp. The validity of physics models is investigated employing ISCDB data. Bayesian model comparison shows differences in the confinement scaling of data subgroups. Theory-based assessment of pure neoclassical transport regimes, however, indicates scalability which is supported by experimental results in specific W7-AS scenarios. Therefore, neoclassical simulations are employed for predictive purposes in W7-X, accounting for effects due to power deposition, plasma profiles and the ambipolar radial electric field. Neoclassical case studies for W7-X are presented as examples for the neoclassical predictions to be considered as an upper limit of plasma performance.

Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6–8.3, β_N values of 2.1–2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5–7, β_N values of 2.3–3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% non-inductive current.

In the 2006 experimental campaign, progress has been made on JET to operate non-inductive scenarios at higher applied powers (31 MW) and density (n_l ∼ 4 × 10¹⁹ m⁻³), with ITER-relevant safety factor (q₉₅ ∼ 5) and plasma shaping, taking advantage of the new divertor capabilities. The extrapolation of the performance using transport modelling benchmarked on the experimental database indicates that the foreseen power upgrade (∼45 MW) will allow the development of non-inductive scenarios where the bootstrap current is maximized together with the fusion yield and not, as in present-day experiments, at its expense. The tools for the long-term JET programme are the new ITER-like ICRH antenna (∼15 MW), an upgrade of the NB power (35 MW/20 s or 17.5 MW/40 s), a new ITER-like first wall, a new pellet injector for edge localized mode control together with improved diagnostic and control capability. Operation with the new wall will set new constraints on non-inductive scenarios that are already addressed experimentally and in the modelling. The fusion performance and driven current that could be reached at high density and power have been estimated using either 0D or 1–1/2D validated transport models. In the high power case (45 MW), the calculations indicate the potential for the operational space of the non-inductive regime to be extended in terms of current (∼2.5 MA) and density (n_l > 5 × 10¹⁹ m⁻³), with high β_N (β_N > 3.0) and a fraction of the bootstrap current within 60–70% at high toroidal field (∼3.5 T).

We report on recent advances in modelling and experiments on resistive wall mode feedback control. The first experimental demonstration of feedback suppression of rotating external kink modes near the ideal wall limit in a tokamak is described [1]. This was achieved using an optimized control system employing a low latency digital controller and directly coupled modular feedback coils. The magnitude of plasma dissipation affecting kink mode behaviour has also been experimentally quantified for the first time using measurements of the radial eigenmode structure of the poloidal field fluctuations associated with the rotating kink mode. New capabilities of the VALEN code [2] are also reported. These include the ability to simulate multiple plasma modes and mode rotation in the model of the feedback control loop. Results from VALEN modelling of resistive wall mode feedback control in ITER are also presented, showing a significant improvement in performance with internal coils. Evidence for a lack of mode rigidity in HBT-EP is given, and plans to address this and other issues related to coil coverage and coil modularity are presented.

In efforts to develop advanced tritium breeders, the effects of additives to lithium titanate (Li₂TiO₃) have been investigated, and good prospects have been obtained by using oxide additives such as TiO₂, CaO and Li₂O. As for the neutron multiplier, the development of a real-size electrode fabrication technique and the characterization of beryllium-based intermetallic compounds such as Be–Ti and Be–V have been performed. Properties of Be–Ti alloys have been found to be better than those of beryllium metal. In particular, steam interaction of a Be–Ti alloy was about 1/1000 as small as that of beryllium metal. These activities have led to bright prospects for the realization of the water-cooled DEMO breeder blanket by application of these advanced materials.

Fishbone instability excited by the supra-thermal circulating electrons in tokamaks is investigated. It is found for the first time that the precession of all the circulating electrons is in the ion diamagnetic drift direction if magnetic shear is neglected. The circulating electrons, which experience both the high field side and the low field side, play a bigger role on the modes than the barely trapped electrons. The analyses show that the mode frequency is close to the precession frequency of circulating electrons comparable with experiment observations. The correlation of the theory with experiments is discussed.

Recent JET and JT-60U results, have underlined the need to look into the measurement requirements of fast particles diagnostics in ITER. Not only alpha particles but all fast ions (He⁴,p,D,T,He³) present in the plasma should be diagnosed. Furthermore, the electron diagnostics, monitoring the deviations from a Maxwellian distribution function, play a central role in the electron cyclotron heating and current drive. In the present ITER measurement requirements only alpha particles and He ions are mentioned: space resolution is δr = a/10 (a is the minor radius), time resolution is δτ = 100 ms and accuracy 10–30%. The new proposed measurement requirements for fast particles are spatial resolution ∼a/20 (10 cm on ITER), time resolution (minimum) ∼100 τ_A (∼100 µs for ITER), density of fast ions: the minority ions could be 4–10% of the plasma density. The candidate diagnostic techniques considered for ITER are amongst others γ-ray spectroscopy, fast ion collective Thomson scattering (CTS) and charge exchange recombination spectroscopy (CXRS). γ-ray spectroscopy detects ions in the energy range 1 < E < 5 MeV and R&D is needed to demonstrate the feasibility of these measurements in the presence of a neutron background. Techniques using ultraviolet spectroscopy of ions such as Krypton are being tested, for the low energy part of spectrum of fast particles. A design study of CTS shows that the requirements are closely met; the system includes a high power gyrotron operating in a region not yet tested in current applications (i.e. using a gyrotron at a frequency below the first ECE harmonic). For CXRS the spatial resolution and accuracy are not achievable for r/a < 0.3 and R&D is needed to assess the minimum figures in terms of accuracy possible for measurements close to the centre. The measurement requirements deduced from the need for detecting the effects of interaction of fast particles with MHD instabilities put forward important new objectives to the candidate diagnostic systems. In particular diagnostics for fast particles in the energy region 200 < E < 1300 keV and diagnostics for escaping fast particles need urgent development.

The establishment of the current profile as in the hybrid scenario is studied under the condition of dominant electron heating in HL-2A. The scenarios with injecting lower hybrid (LH) and electron cyclotron (EC) waves are under numerical study. Carefully adjusting the position of non-inductive current driven by two groups of gyrotron, an optimized q-profile was obtained with q_a = 3.78 and a weak shear region extending to ρ ∼ 0.45 (where ρ is the square-root of toroidal flux normalized to its value at the plasma boundary) in low-density discharges of . When 0.5 MW LH power in the current drive mode and 0.95 MW EC power mainly for plasma heating are used to control the current profile, a hybrid discharge scenario with a weak magnetic shear region extended to ρ = 0.6 and q_a = 3.21 is established by controlling the EC absorption position. The mechanism of the LH wave absorption in the HL-2A plasma causes interplay of the distribution of the LH driven current with the modification of the plasma configuration, which constitutes non-linearity in the LH wave deposition. Due to the non-linearity the LH wave deposition position changes spontaneously or oscillates. The oscillatory behaviour caused by the non-linear effect of the LH wave deposition is analysed.

For the first time, scalings for density peaking in tokamaks are obtained from a database consisting of observations from two devices, ASDEX Upgrade and JET. The investigation relies on an inversion method for the interferometer signals which grants consistent reconstructions despite differences in interferometer geometries. By combining observations from these devices, correlations between physics parameters investigated for their role in determining density peaking are reduced. Multiple regression analyses show that in the combined database collisionality is the most relevant parameter. The particle source provided by neutral beam injection provides a contribution to the peaking, which, although not negligible, is not large enough to explain the whole observed variation of density peaking. The device size, introduced as an alias for possible systematic differences between the devices not captured by the regression parameters, is found to play only a small role in regressions which include collisionality. Device size becomes relevant in scalings which exclude collisionality and include the ratio of the density to the Greenwald density limit. This indicates that density peaking is more likely to be a function of collisionality rather than of the fraction of the density limit. All the scalings which include collisionality in the regression variables predict a peaked density profile for the ITER standard scenario.

In a tokamak, misalignments of field coils and coil connections, introduced in the construction of the device, lead to small asymmetries in magnetic field known as error fields. Even at small levels these can stop the plasma rotating, driving resistive tearing modes that lead to plasma termination. Recently, a set of four error field correction coils has been installed on the MAST spherical tokamak. These coils have been used to correct for the error field, allowing MAST to operate in previously inaccessible regimes. With error field correction, MAST can operate at densities at least 30% lower than previously attained. The error field coils have also been used to apply large error fields to the plasma to explore the phenomenon of mode locking. It was found that the locked mode threshold scales with line averaged electron density, safety factor and toroidal field in a similar way to that found from conventional aspect ratio tokamaks. This scaling law has been used to estimate the locked mode threshold for a proposed power plant and a component test facility based on the spherical tokamak concept.

An expression for the extrapolated fusion gain G = P_fusion/5P_heat (P_fusion being the total fusion power and P_heat the total heating power) of ITER in terms of the confinement improvement factor (H) and the normalized beta (β_N) is derived in this paper. It is shown that an increase in normalized beta can be expected to have a negative or neutral influence on G depending on the chosen confinement scaling law. Figures of merit like should be used with care, since large values of this quantity do not guarantee high values of G, and might not be attainable with the heating power installed on ITER.

A formation and confinement experiment for fast ions is performed using the ion cyclotron range of frequencies (ICRF) minority heating scheme with a proton minority and a deuteron majority in Heliotron J, a low-shear helical-axis heliotron. The effect of the magnetic configuration on the fast-ion confinement is one of the most important issues in helical devices. In this paper, the effect of bumpiness, one of the Fourier components of the field strength, on the trapped fast-ion confinement is clarified by using ICRF minority heating. The role of the bumpiness is a key issue for the design principle of the magnetic field of Heliotron J, where the particle confinement is controlled by the bumpiness. Here, the bumpiness or the bumpy ripple is defined by the Fourier harmonic ratio ε_b ≡ B₀₄/B₀₀ in (Mizuuchi T et al 2006 Fusion Sci. Technol.50 352), where B₀₄ is the bumpy component and B₀₀ is the averaged magnetic field strength. The proper bumpiness causes deeply trapped particles to be confined in the small grad-B region. High-energy ions are produced up to 10 keV by injecting an ICRF pulse into an electron cyclotron heating target plasma where ion temperature at the centre T_i(0) = 0.2 keV, electron temperature at the centre T_e(0) = 0.8 keV and line-averaged electron density . For the study of the configuration dependence of the fast particle confinement, three configurations are selected; the bumpy ripples are 0.01, 0.06 and 0.15 at the normalized minor radius ρ = 0.67. The measured tail temperatures by using a charge-exchange neutral energy analyser are 1.10 keV, 0.88 keV and 0.50 keV for the ripples of 0.15, 0.06 and 0.01, respectively. The heating efficiency of the bulk ion is also better in the high bumpy case. A Monte-Carlo analysis also indicates good confinement of the high energy ions in the high bumpy case although the difference is not very large compared with the experiment.

The first plasma has been achieved successfully in the Experimental Advanced Superconducting Tokamak (EAST). Boronization by the glow discharge (GDC) method was studied in experiments. The plasma performance was obviously improved by GDC boronization. Extension of the operational region and improvement in the plasma performance were obtained. Sawtooth discharges were observed by means of soft x-ray signals, electron cyclotron emission signals and line averaged electron density after boronization. Lower q_a and more stable operation were also achieved following GDC boronization. The plasma current ramp-up rate was also improved as a result of decreased impurity content and low averaged loop voltage due to boronization. PLEASE NOTE: THERE HAS BEEN A RETRACTION PUBLISHED FOR THIS ARTICLE.

The effects of a transient heat load on tungsten damaged by helium plasma irradiation have been investigated using a ruby laser with long pulse duration in the divertor simulator NAGDIS-II (Takamura et al2002 Plasma Sources Sci. Technol.11 A42). The pulse width of the ruby laser was ∼0.6 ms, which is close to that of the expected heat load accompanied by type-I edge localized modes (ELMs) in ITER operation. Helium holes/bubbles, which were formed in the surface region of powder metallurgy tungsten due to the exposure to the helium plasma, disappeared after the laser pulse irradiation to the tungsten surface with sufficient pulse energy. The results indicated that the transient heat loads similar to those expected by ELMs will mitigate damages such as bubbles and holes produced by helium irradiation. When a vacuum plasma sprayed tungsten coating on graphite was exposed to the helium plasma, the surface was covered with arborescent nanostructured tungsten containing many helium bubbles inside the structure. Melting traces were found on the surface after the laser pulses irradiated the surface even though the pulse energy was lower than that for melting bulk tungsten. A numerical temperature calculation of the sample suggested that the effective thermal conductivity near the surface dramatically decreased by several orders of magnitude due to the formation of nanostructured tungsten.

Interpretative transport has been used to revisit the global scalings of TJ-II plasmas with electron cyclotron resonance heating (ECH), from a local perspective. Density, rotational transform and ECH power scans were analysed based mainly upon Thomson scattering data (electron density and temperature) in steady state discharges. The thermal diffusivity χ_e is estimated assuming pure diffusion and negligible convective heat fluxes in a set of 201 discharges. The density scan indicates that for normalized minor radii ρ ≲ 0.4 there is no significant change of χ_e with density in the range studied , while in the density gradient region, χ_e decreases with line density. In the rotational transform scan, there is an overall tendency of χ_e to decrease with increasing rotational transform, but we also find that the presence of a low order rational value of the rotational transform is accompanied by a local lowering of χ_e. Finally, in the ECH power scan, χ_e is found to have an overall increment in 0.2 < ρ < 0.6 when Q_ECH increases from 200 to 400 kW, although it is less significant in the density gradient region.

Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S ≡ q₉₅ (I_p/aB_t)) of ∼41 (MA m⁻¹ T⁻¹) simultaneous with a record plasma elongation of κ ≡ b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance. Plasmas with high shape factor have been sustained for pulse lengths which correspond to τ_pulse = 1.6s ∼ 50 τ_E ∼ 5 τ_CR, where τ_CR is the current relaxation time and τ_E is the energy confinement time. Plasmas with higher β_t ∼ 20% have been sustained for τ_pulse = 1.2s ∼ 25 τ_E ∼ 3 τ_CR with non-inductive current fractions f_NI ∼ 50%, with ∼40% pressure driven current and ∼10% neutral beam driven current. An interesting feature of these discharges is the observation that the central value of the safety factor q(0) remains elevated and constant for several current diffusion times without sawteeth, similar to the 'hybrid mode'. Calculations of the profiles of inductive and non-inductive current are compared with measurements of the total current profile and are shown to be in quantitative agreement. Results are shown from experiments which investigate the applicability of high harmonic fast waves (HHFWs) and electron Bernstein waves (EBWs) as current drive and heating sources, and the possibility of LHCD for future ST devices is raised. A calculated scenario which provides 100% non-inductive current drive is described. NSTX operates with peak divertor heat fluxes which are in the same range as those expected for the ITER device, i.e. with . High triangularity, high elongation plasmas on NSTX have been demonstrated to have reduced peak heat flux to the divertor plates to <3 MW m⁻².

The gyrokinetic-Vlasov simulation code (GKV code) is applied to zonal flows and the ion temperature gradient (ITG) turbulence in helical systems with L = 2 and M = 10 like the Large Helical Device (where L and M denote poloidal and toroidal periodicities of the main helical component of the confinement field, respectively) for the standard and inward-shifted model configurations. Because of the slower radial drift motion of helical-ripple-trapped particles, the inward-shifted case provides a higher zonal-flow response than that in the standard model with smaller side-band helical field components. The nonlinear GKV simulations show that the ITG turbulent transport in the inward-shifted model, which has larger growth rates of the ITG stability, is regulated by the zonal flows to a level comparable to the standard case.

In parallel to the ITER design process and in close cooperation with the designers a fusion-specific safety approach was developed and implemented. Detailed safety assessments have been performed and documented in the ITER Generic Site Safety Report (GSSR). Following the decision on ITER construction in France, results from the GSSR and from on-going safety-related activities tailored to the Cadarache site and the French licensing process are now being used to write the ITER Preliminary Safety Analysis Report.

In the most recent European fusion power plant conceptual study (PPCS) inherent fusion favourable features have been exploited, by appropriate design and choice of materials, to provide major safety and environmental advantages. The study focused on five power plant models, which are illustrative of a wider spectrum of possibilities. These span a range from relatively near-term concepts, based on limited technology and plasma physics extrapolations, to a more advanced conception. All five PPCS plant models differ substantially in their plasma physics, blanket and divertor technology, size, fusion power and materials compositions, and these differences lead to differences in economic performance and in the details of safety and environmental impacts.

This paper uses the quite detailed information available from ITER safety documents and highlights the differences between ITER and future fusion power plants. The main areas investigated are releases and doses during normal operation and under accidental conditions, occupational radiation exposure and optimization and waste management, including recycling and/or final disposal in repositories.

Due to an error, an incorrect version of this paper was published in issue 7. For the convenience of the reader we have included the correct full article below rather than a list of changes.