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The JT-60SA superconducting tokamak is now under construction toward the first plasma in March 2019 as a joint project between the Broader Approach (BA) Satellite Tokamak Programme between Japan and Europe, and the Japanese national programme. The JT-60SA mission is to contribute to early realization of fusion energy by supporting ITER and by complementing ITER in resolving key physics and engineering issues for DEMO reactors. Before procurements of the major components, some R&Ds for key techniques were performed. By November 2014, 25 procurement arrangements (PAs) have been launched (JA: 14 PAs, EU: 11 PAs) covering 88% of the total cost of the BA Satellite Tokamak Programme, and the main components have entered the manufacturing stage. In addition, the JT-60SA tokamak assembly started since January 2013. This paper summarizes the recent progress of the JT-60SA project.

Spatio-temporal measurements of the impurity ion temperature, density, and rotation profiles around the plasma edge region have been made in the JT-60U tokamak, allowing the determination of radial electric field, Er, with the key dimensionless parameter (poloidal Mach number, Upm) at the L-H transition in a number of operational regimes. We found that there is variation in the L-H transition in terms of its time-scale; not only “hard” type transition with a faster time-scale than that seen in the plasma transport (as represented by an energy confinement time, τE) as seen in the many conventional tokamaks, but also “soft” one with a slow time-scale (≈τE) is possible solution, including a complex multi-stage Er transition in the later H-phase. The most important point is that the critical condition for the L-H transition predicted by ion-orbit loss model could be applicable only for “hard” transition (occurred at Upm ≥ 1), and not necessary for “slow” one (occurred even at Upm < 1). Characteristics of the turbulent density fluctuation with the frequency range of 100 kHz at the plasma edge region, in addition to a uniform toroidal MHD oscillation (i.e., n = 0), during ELM-free H-phase are also reported. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Present status of the JT-60SA (JT-60 Super Advanced) project, implemented jointly by Europe and Japan since 2007, is described. The design of the main tokamak components was completed in late 2008, and all the scientific missions are preserved to contribute to ITER and DEMO reactors. The construction of the JT-60SA has begun with procurement activities for the superconducting magnet systems, vacuum vessel, in-vessel components and other components under the relevant procurement arrangements between the implementing agencies of JAEA (Japan Atomic Energy Agency) in Japan and Fusion for Energy in Europe. Designs and developments of the auxiliary heating systems for JT-60SA have been progressing at JAEA so as to provide the total injection power of 41 MW for 100 s.

The JT-60SA (Super Advanced) project is a joint project of the ITER Satellite Tokamak program and the National Centralized Tokamak program in Japan with missions of supporting ITER, complementing ITER and exploring advanced issues toward DEMO. JT-60SA is a tokamak with superconducting coils, equipped with a poloidal field coil system with wide plasma shape controllability, upper and lower divertors with different shapes, NBI and ECRF with heating power 41 MW and various heating methods, in-vessel coils for suppressing MHD instabilities. With these functions, possibilities of producing ELMy H-mode with improved confinement, full non-inductive current drive of high beta plasmas (β N =3. 7 at Ip=3.5 MA, β N =4.4 at Ip=2.4 MA) and break-even class plasmas necessary for accomplishing the mission have been confirmed. The engineering design of JT-60SA is being done taking large annual neutron production into account. Double skin walls filled with borated water or boron doped concrete are employed for the vacuum vessel and cryostat, respectively, for neutron shield. Divertors structures and first walls are being designed so as to be changed with remote handling systems in the high radiation circumference.

In the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U), a high- β p ELMy H-mode (high-poloidal-beta high-confinement-mode with edge localized mode) plasma was sustained with β N ∼2.7 for 7.4 s. Real-time neoclassical tearing mode (NTM) stabilization system was established and effective NTM suppression by early electron cyclotron (EC) wave injection was demonstrated. High fusion triple product of n i (0)τ E T i (0)=3.1×10 20 keV⋅s⋅m −3 was achieved using the negative-ion based neutral beam current drive with β N ∼2.5 and the bootstrap current fraction f BS ∼50%. In a hot electron regime, a high electron cyclotron current drive efficiency of 4.2×10 18 A/W/m 2 was achieved at T e ∼21 keV . Innovative current start-up scenario produced a current hole plasma with a very high f BS ∼90%. No accumulation of helium and carbon impurities was observed for internal transport barrier (ITB) plasmas. While argon impurity was accumulated, EC injection effectively exhausted it across ITB. In a regi...

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.

Two types of advanced tokamak scenarios, namely, high βp H-mode and reversed shear (RS) mode, are developed and studied extensively in Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ide and the JT-60 Team, Phys. Plasmas 7, 1927 (2000)]. These plasmas have both an internal transport barrier (ITB) and an edge transport barrier (ETB). The goal of the recent studies was therefore to get a deeper insight into the physics of ITB and ETB and to increase the discharge performance under a controlled manner based on the physics understanding. In the studies of ETB in an extended high triangularity operation with pellet injection from the high-field side, we have clarified the linkage between main parameters determining the pedestal structure. In the studies of type I edge localized modes (ELMs), simultaneous fast ELM measurements revealed four different phases in an ELM event. It was also established that the density collapse is poloidally asymmetric and is localized at the low-field side. In the studies of ITB, the response of the ion thermal diffusivity to the radial electric field (Er) shear and the effect of electron heating on ITB have been investigated. The importance of nonlocality and the effective Er shear are discussed. In the case of the RS plasma with a strongly reversed current profile, it has been found that there is no plasma current in the central region (called the “current hole”), and it is impossible to drive current in either the positive or negative direction by an electron cyclotron wave inside the current hole. Based on the new understanding of ETB and ITB, the control of the transport barriers is also discussed briefly.

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state oper...

Termination of the runaway electron current generated during plasma disruptions is found in JT-60U during simulated vertical plasma displacement events where the safety factor at the plasma surface qs decreases. For all discharges with runaway electron generation, the runaway current disappears for qs≥2 with the appearance of spikes in the magnetic fluctuations. The growth rate of the spikes in the magnetic fluctuations decreases by an order of magnitude during the termination of runaway current. Corresponding to the loss of runaway electrons by magnetic fluctuations, heat flux pulses are measured at the inner divertor plates, which indicates interaction of the wall with the runaway electrons. The halo current during runaway termination is small and increases after runaway termination with a dominant toroidal mode of n = 1.

With the main aim of providing the physics basis for ITER and steady state tokamak reactors, JT-60U has been optimizing operational concepts and extending discharge regimes towards simultaneous sustainment of high confinement, high βN, high bootstrap fraction, full non-inductive current drive, and efficient heat and particle exhaust utilizing a variety of heating, current drive, torque input and particle control capabilities. In the two advanced operation regimes (reversed magnetic shear (RS mode) and weak magnetic shear (high βp mode)), ELMy H mode discharges characterized by both internal and edge transport barriers and by high bootstrap current fraction fBS have been sustained near the steady state current profile solutions under full non-inductive current drive with appropriate driven current profiles (high βp: HHy2 ≈ 1.4 and βN ≈ 2.5 with negative ion based neutral beams (N-NBs); RS: HHy2 ≈ 2.2 and βN ≈ 2 with fBS ≈ 80%). Multiple pellet injection has extended the density region with high confinement. These operational modes have been extended to the reactor relevant regime with small values of collisionality and normalized gyroradius, and Te~Ti. In the RS regime, QDTeq = 0.5 has been sustained for 0.8 s. Stability has been improved in the high βp mode regime by suppression of the neoclassical tearing mode with local ECCD and in the RS mode regime by wall stabilization. The internal transport barrier structure has been controlled by toroidal rotation profile modification, and transport studies have revealed a semiglobal dynamics of the internal transport barrier structure. Simultaneous pumping at the inner and the outer divertor leg has enhanced He exhaust by ~40%. Argon puff experiments have improved confinement at high density with detached divertor owing to high pedestal temperature Tiped. In H modes, the core confinement degraded with decreasing Tiped, suggesting stiff core profiles. An operational region of grassy ELMs with small divertor heat load has been established at high triangularity, high q95 and high βp. A record value of the neutral beam current drive efficiency of 1.55 × 1019A m-2 W-1 has been demonstrated using N-NB. Abrupt large amplitude events causing a drop in neutron emission have been discovered with frequency inside the TAE gap. Disruption studies have clarified that runaway current is terminated by MHD fluctuations when the surface q decreases to 2-3.

Progress on NNB technology enables to demonstrate the N-NB driven current of around 1 MA with the efficiency of 0.155×1020nT−2A/W. A newly installed 110GHz ECH system has demonstrated efficient cen...

The JT-60U [Y. Kusama and the JT-60 team, Phys. Plasmas 6, 1935 (1999)] high performance plasmas, such as high poloidal-beta high-confinement mode (high βp H-mode) and reversed magnetic shear (RS) plasmas, have been optimized towards a regime of steady state operation concerning high confinement, high β and high bootstrap current fraction which are requirements of a steady state fusion plasma. What is essential to keep improved confinement in these plasmas is to maintain the internal and/or the edge transport barriers (ITB and/or ETB). A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In JT-60U, RS plasmas have been optimized not only for the high performance but also for reactor relevant steady state research. In 1999, as the result of the optimization, the equivalent deuterium–tritium (D–T) fusion gain (QDTeq) of 0.5 was sustained for 0.8 s, which is roughly equal...

This paper describes the latest achievements of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [H. Kimura and the JT-60 Team, Phys. Plasmas 3, 1943 (1996)] for the establishment of the physics and technology basis for steady-state, fully noninductive current-drive plasmas with high fusion performance. Recent results are highlighted by (1) high performance reversed-magnetic-shear discharges [an idealized equivalent QDT of 1.05, a confinement-enhancement factor (H factor) of 3.23, and a normalized beta value (βN) of 1.88 at Ip=2.8 MA/Pabs=17 MW]; the first observation of thermal transport barriers both for electrons and ions, (2) the first injection of negative-ion-based neutral deuterium beams in the world for studies of current drive, heating, and high-energy particle behavior (achieved so far: 2.5 MW/400 keV, design values: 10 MW/500 keV); high neutralization efficiency of 60% at 370 keV; high current drive efficiency of ηCD=8×1018 m−2 A/W, (3) improved giant edge-localized-mode (...

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e.g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.

This paper summarizes major plasma physics obtained during the period of JT‐60 operation for 23 years with special emphasis on research towards steady‐state operation of tokamak. Topics included are observation of large bootstrap current fraction, negative shear scenario, demonstration of efficient beam and EC current drive, discovery of current hole, stabilization of the RWM, discovery of ITB (internal transport barrier).

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

Beam acceleration by heating in the second-harmonic ion cyclotron range of frequency (ICRF) in combination with heating by hydrogen neutral beam injection (NBI) was investigated in the JT-60 tokamak. The energy spectra of the accelerated fast ions were measured by a charge exchange neutral energy analyser whose line of sight was intersected by specific beam lines in the plasma core in order to obtain the required information. The dependences of the tail ion temperature on various parameters (electron density, NBI power and the toroidal phasing of the antenna) were examined. Optimized conditions for beam acceleration were evaluated. For combined ICRF heating and NBI heating, an incremental energy confinement time of 210 ms was achieved, which was three times higher than that obtained with NBI heating alone or with ICRF heating alone. This improvement of the energy confinement during combined NBI and ICRF heating can be explained by a build-up of the fast ions accelerated by the ICRF wave. The scaling of the incremental energy confinement time during combined ICRF+NBI heating was obtained.

The energy confinement in high power lower hybrid current driven (LHCD) plasmas has been studied in the JT-60 tokamak. At a plasma current of 1 MA, the diamagnetically estimated energy confinement time in LHCD plasmas has almost the same value as the confinement time in ohmically heated plasmas at e ~ 1.0 × 1019 m−3. The confinement time of high power LHCD plasmas (PLH ≤ 4.5 MW) is longer than that of neutral beam heated plasmas at the same power level. The energy confinement time depends on the LHCD power and on the electron density, but is independent of the plasma current, with α + β ~ −0.3.

Plasma current (Ip) initiation and ramp‐up utilizing induction by poloidal field (PF) coils located on the outboard side of the torus were demonstrated in both conventional aspect ratio tokamak (100 kA in JT‐60U) and spherical tokamak (10 kA in TST‐2). In the presence of sufficient source of plasma by RF ionization (1 MW in JT‐60U, 100 kW in TST‐2), Ip formation is possible without the existence of a PF null, with the initial vertical field (Bv) in the direction opposite to that required for equilibrium. Utilization of the extra flux swing provided by the PF coils before Bv switches sign is important in arriving at a high enough Ip that matches Bv at the end of PF coil current ramp. In JT‐60U, recharging of the OH coil was observed with only perpendicular and counter NB injection, suggesting overdrive by bootstrap current. In TST‐2, a quasi‐steady‐state Ip of 4 kA was formed and sustained by RF power (PRF) alone, without the use of induction. The plasma current centroid of this type of plasma is located on the outboard side. The dependences of Ip on PRF and Bv were found to be weak. The results presented here offer encouragement to CS‐less operation of a tokamak fusion reactor, which should have a large impact on the economic competitiveness.

A central electron temperature of 23 keV was obtained by a combination of ECH and ECCD. A highly localized current density profile was obtained by central ECCD, indicating a diffusion coefficient for resonant electrons of at most 1 m2/s. A comparison of HFS and LFS off‐axis ECCD confirmed the theoretically expected trapped particle effect on the current drive efficiency. LHCD was used to broaden the qmin radius and the ITB foot radius. N‐NBCD was used to shrink the current hole radius. An integrated scenario consisting of (1) a novel plasma start‐up method using RF plasma production and induction by vertical field and shaping coils, (2) a noninductive ramp‐up stage, and (3) a transition to a high‐density, bootstrap‐dominated, high‐confinement plasma, without the use of the OH solenoid, has been demonstrated. The plasma created by this technique had both internal and edge transport barriers, with βp = 3.6 (eβp=1), βN = 1.6, HH98y2 = 1.6 and fBS 90% at Ip = 0.6 MA and ne = 0.5nGW.

Modulation charge eXchange recombination spectroscopy measurements with high spatial and temporal resolution have made the evaluation of the toroidal plasma flow of fully stripped carbon impurity ions ( VϕC6+) in the JT-60U tokamak peripheral region (including, in particular, the separatrix) possible with a better signal-to-noise ratio. By comparing co- and counter-neutral beam injection discharges experimentally, we have identified the boundary condition of VϕC6+ and radial electric field shear ( ∇Er) imposed at the separatrix in high confinement (H-mode) plasmas with edge localized modes (ELMs). The VϕC6+ value at the separatrix is not fixed at zero but varies with the momentum input direction. On the other hand, the ∇Er value is nearly zero (or very weakly positive) at the separatrix. Furthermore, the edge localized mode perturbation does not appear to affect both VϕC6+ and ∇Er values at the separatrix as strongly as that in the pedestal region. The above experimental findings based on the precise edge...

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified ca...

The duration of advanced tokamak plasma operation in JT-60 Upgrade (JT-60U) [A. Kitsunezaki et al., Fusion Sci. Technol. 42, 179 (2002)] has been extended on long time scales exceeding the current diffusion time ( τ R ) and close to the wall saturation time. A very high bootstrap current fraction ( f BS ) of 75% was maintained for 7.4 s ( 2.7 τ R ) , while a normalized beta of 2.5 was maintained for 15.5 s ( ∼ 9.5 τ R ) . The current profile reaches stationary conditions in ∼ τ R for the small f BS regime, while a longer time is required for the large f BS regime. A plasma with a weak shear q profile, similar to requirements for the ITER steady-state operational scenario, was successfully maintained. The particle inventory in the wall was observed to saturate in repeated long-pulse ( ∼ 30 s ) H -mode discharges with edge localized modes. The analysis of neutral particles in the scrape-of-layer plasma indicates different time scales are involved for the wall saturation. Ferritic steel tiles have been insta...

Results are presented of the first Lower Hybrid Current Drive (LHCD) experiments in JT-60. 2 MA of RF driven current is successfully produced for the first time in a reactor grade tokamak. The magnetic divertor works quite well in eliminating the impurities released by the current carrying fast electrons which have allowed the generation of the reactor relevant RF current in a very low density plasma. The efficiency which is defined as ηCD = eRIRF/PLH(1019 m−3 AW−1), reaches values of 0.8 to 1.7. NBI heating enhances the current drive efficiency by a factor of 1.5, and LHCD improves the confinement time of high power NBI heated plasma. The key to confinement improvement is found to be the active control of the current profile by LHCD.

The characteristics of ohmically and high power neutral beam heated (PNBI ~ 20 MW) hydrogen divertor plasmas in JT-60 are analysed on the basis of a plasma fluid model. The calculated radiation loss of the divertor plasma and the density dependence of the neutral pressure in the divertor chamber are qualitatively in good agreement with the experimental results. The divertor radiation loss in ohmically heated discharges can be explained by hydrogen line radiation and impurity radiation. The higher neutral pressure in neutral beam heated discharges can be explained by the decrease of the global particle confinement time in comparison with that in ohmically heated discharges.

This paper investigates the current buildup just after the breakdown, and it becomes clear that stable plasma initiation can be obtained by the precise control of particles contained in the graphite wall using clean-up shots with/without gas puff.

The characteristics of particle confinement and recycling of ohmically and neutral beam (NB) heated hydrogen plasmas in the JT-60 tokamak are investigated by using the global particle balance equation. The total number of ionization events is estimated by employing a simple formula for the Hα radiation profile, and the gas fuelling efficiency is determined from the time derivatives of the electron density before and after gas feed termination. The global particle confinement time in the Ohmic heating divertor plasma decreases nearly inversely proportionally to the electron density, because of the enhancement of recycling in the plasma periphery. NB injection deteriorates the particle confinement time by a factor of about two. The effective particle confinement time closely related to the density control increases proportionally to the electron density in the Ohmic plasma. In contrast, the effective particle confinement time of the NB heated plasma is independent of the electron density although it decreases with the NB injection power. These results show that the plasma density can be controlled by the gas puff in both ohmically and NB heated hydrogen discharges.

Characteristics are presented of the energy confinement time resulting from neutral hydrogen beam heating of JT-60 tokamak plasmas with a plasma current, Ip, of 1-2 MA, a line averaged electron density, e, of 1.5-7 × 1019 m−3 and an absorbed beam power, Pabs, up to 20 MW. The plasma energy content follows an offset linear relation to the absorbed power. The incremental energy confinement time for thermal components, , is almost independent of Ip and e at 60 ms, while the incremental energy confinement time for the total plasma energy, , decreases with plasma density. The value of in JT-60 coincides with the prediction of the Shimomura-Odajima scaling of L-mode discharges.

Lower hybrid current drive (LHCD) experiments on JT-60 have been carried out in parameter ;e = (0.6−1.6)×1019 m−3, Ip = 0.5−1.5 MA, BT = 2.7−4.5 T, PLH < 2.5 MW, f = 1.75 and 2.0 GHz, with and without NB heating (PNB < 6 MW), limiter and lower side single null divertor configurations, and narrow and wide spectra. The current drive efficiency increases with the volume averaged electron temperature Te and with decreasing Zeff and a high efficiency of ηCD ≡ eRPIRF/PLH ≤ 2.65 × 1019 m−2AW−1 was achieved by LHCD alone. The experimental data set has shown that the current drive efficiency is consistent with the scaling of ηCD ~ 12 Te/(5+Zeff)×l019 m−2AW−1. Low N1 waves with a narrow spectrum have a higher current drive efficiency than higher N1 waves with a wide spectrum insofar as launched waves satisfy the accessibility

Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).