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TIMESTAMPS

The JT-60U tokamak project has addressed major physics and technological issues for burning plasma development in ITER and for steady-state high-b operation development towards JT-60SA and DEMO reactors. The fusion research has to establish an efficient control of the self-regulating plasmas for achieving a high integrated performance. Toward this goal, in particular, the mechanisms determining the plasma rotation profile and effects of the rotation on transport and stability are the central research issues. Reduction of the toroidal field ripple by installing ferritic steel tiles in JT-60U (installed in 2005) reduced the fast ion losses and resultant counter plasma rotation drive. By combining this newly achieved freedom of rotation with co-, counter- and perpendicular NBs, JT-60U has been promoting an integrated research project focusing on the plasma rotation covering transport, stability, pedestal, and steady state operation.

Application of the real time current profile control was extended to higher β_N plasmas. Control of the minimum in the q profile (q_min) was successfully demonstrated [1.1-1]. Pursue of a plasma with high bootstrap current fraction (f_BS) had been continued [1.1-2, 1.1-3]. Duration of sustainment of high f_BS, about 70%, was extended to about 8 s. A plasma sustained fully by the bootstrap current was confirmed by the loop voltage profile analysis. Current ramp up solely by the bootstrap current was also demonstrated.

Real time current profile control by using both real time evaluation of the q profile from the MSE diagnostics and real time control on the LHCD system had been extended to higher performance plasmas. It was demonstrated that in a plasma with higher β_N. The control was applied to high-β_p mode plasma (I_p = 0.8 MA, B_t = 2.5 T, q₉₅ = 5.9) having initial q_min of about 1.2. The plasma having the normalized beta β_N of about 1.3-1.6 was by 11 MW of NB heating. Real time control of the q_min was successfully demonstrated in the plasma as shown in Fig.I.1.1-1. The reference q_min was increased from 1.3 to 1.7 taking 3 s, the actual q_min followed the reference and reached 1.7. When the q_min was controlled at the fixed value of 1.7, the central electron temperature (r/a < 0.3), the line averaged electron density and stored energy started increasing, which shows formation of an internal transport barrier (ITB). The application of the real-time control changed the q profile and lead to the change in confinement. Due to the resulting increase in the electron temperature and the pressure, central Ohmic current and bootstrap current increased, leading to decrease in q_min. However, the decrease in q_min was recovered by the increase in P_LH by the control, here P_LH is the LH power. The real-time q_min control was demonstrated in such a self-regulated plasma, where the pressure and the current profiles are linked through the bootstrap current.

Capability of off-axis NBCD had been assessed, in plasmas with I_p = 0.8 MA and 1.2 MA at B_t = 3.8 T [1.1-1]. In both cases, a well localized NB driven current profile was confirmed for the first time. The total amount of the driven current estimated from the MSE measurement almost agreed with calculations by the ACCOME code in both cases. However, the measured j_NB (ρ) was more off-axis that that in the calculations.

In a previous campaign, sustainment of high f_BS of 75% for 7.4 s was demonstrated [1.1-4]. It had been tried to extend this period, and finally duration of high f_BS sustainment, though f_BS is a litter lower in this case (~70%), was extended to 8 s. Furthermore, a self-sustained state driven by the bootstrap current was achieved. A loop voltage profile of nearly zero or slightly negative across the cross section was maintained for 0.2 s with a constant OH primary current. In this discharge, of the total I_p of 543 kA, the calculated beam driven current was -35 kA and the inductive current was -5 kA. In the newly implemented constant plasma surface flux feedback mode, which ensures no net flux input to the plasma, a steady I_p ramp-up at a rate of 10 kA/s was achieved for 0.5 s. These results provide the evidence of bootstrap overdrive.

References
1.1-1 Suzuki, T., et al., Proc. 21^st Int. Conf. on Fusion Energy 2006 (Chengdu, 2006) CD-ROM file EX/6-4.
1.1-2 Sakamoto, Y., et al., Proc. 21^st Int. Conf. on Fusion Energy 2006 (Chengdu, 2006) CD-ROM file EX/P1-10.
1.1-3 Takase, Y., et al., Proc. 21^st Int. Conf. on Fusion Energy 2006 (Chengdu, 2006) CD-ROM file EX/1-4.
1.1-4 Sakamoto, Y., et al., Nucl. Fusion 45, 574 (2005).

The large bootstrap current fraction of f_BS > 0.7 is required in a steady-state tokamak fusion reactor to reduce a circulating power for non-inductive current drivers. In the large f_BS plasma, the current and pressure profiles are strongly linked to each other. A control scheme for such self-regulating large f_BS plasmas with the strong parameter linkage has been developed in the reversed shear (RS) plasmas with f_BS ~ 0.7 [1.2-1]. The large f_BS RS plasma was frequently terminated by the disruption or the collapse when the minimum value of the safety factor (q_min) passed through integer values during the current profile relaxation towards a steady-state profile. In order to avoid the disruption or the collapse, the pressure gradient at the internal transport barrier (ITB) should decrease at q_min ~ integer. It has been demonstrated that the steep pressure gradient at the ITB can be reduced by the control of the toroidal rotation profile [1.2-2]. However, it was difficult to adjust the timing of the rotation control using the pre-programmed setting, because the timing of q_min ~ integer depends on the time evolution of the plasma current profile. The time evolution of the plasma current profile is affected by an initial current profile, an onset of ITB formation, an ITB structure and so on. In order to establish the effective pressure profile control at q_min ~ integer through the rotation control, a real time control logic was newly installed based on the real time calculation of q_min using MSE diagnostics [1.2-3]. In this logic, the counter (ctr) neutral beam injection (NBI) is stopped automatically when q_min is close to the integer value.
In order to optimize the control logic, the response of q_min to the change in the ion temperature gradient at the ITB was first investigated. The strong response was observed in the large f_BS (~ 0.7) plasma compared with the f_BS ~ 0.5 plasma when the ion temperature gradient at the ITB was changed by controlling the toroidal rotation as shown in Fig. I.1.2-1. The strong response indicates the strong linkage between the pressure and current profiles in the large f_BS plasma. Based on this result, the duration for the stop of ctr-NBI was set to be 0.8 s. Note that q_min was not measured during the stop of the ctr-NBI, because the ctr-NBI is diagnostic beam for MSE. Therefore, the duration was set as a pre-programmed value. In the case of the long duration, the degraded ITB was not recovered due to the large change of current profile even when the ctr-NBI was injected again after qmin passed through the integer value. In the case of the short duration, the changes in the pressure gradient at the ITB and the current profile were not sufficient to avoid the disruption and the collapse.
By optimizing this new real time control scheme, the pressure profile was controlled with the good reproducibility just before q_min ~ integer in which the disruption or the collapse occurred frequently. As the result, the weak RS plasma with f_BS ~ 0.7 was successfully sustained for ~ 8 s, which corresponds to 3τ_R. Here, τ_R is the current diffusion time. Typical waveforms of the long sustainment of the RS plasma with the large f_BS using the newly installed real time control logic are shown in Fig. I.1.2-2 (a), where I_p = 0.8 MA, B_T = 3.4 T, q₉₅ ~ 8.5, κ = 1.6, δ = 0.4. Using the feedback control of the stored energy by the perpendicular NBIs, β_N ~ 1.5 (β_p ~ 2.1) was maintained. In this plasma, ctr-NBI was turned off for 0.8 s when q_min was in the range of 3.7 - 4.1 as setting. The value of q_min decreased in time and was below 4.1 at t ~ 7 s as shown in this figure. Then ctr-NBI turned off for 0.8 s. During this phase, the toroidal rotation started to change towards the co-direction, and then the electron pressure at the ITB was reduced and recovered as shown in the figure. Behavior of the ion temperature was similar to that of the electron pressure profile. The reduction of the pressure gradient at the ITB in this period led to the avoidance of the collapse at q_min ~ 4. High confinement of H₈₉ ~ 2.6 (HH_98y2 ~ 1.8) was obtained in this plasma, and f_BS ~ 0.7 was sustained for ~ 8 s.

In the reversed shear plasma, the suppression of long-spatial-scale density fluctuations with a wave number of the order of 1 cm^-1 was observed in the ITB a region, after a cold pulse was induced by the shallow pellet injection at the plasma edge. Particle and power balance analysis with fully recovered temperature profiles after the cold pulse propagation indicated that the particle and the ion heat transport is coupled with the long-spatial-scale density fluctuations, while the electron heat transport is not coupled with them [1.2-4]. Transient response of the electron heat transport during the cold pulse propagation was investigated for better understanding of the relation between the electron heat transport and long-spatial-scale density fluctuations [1.2-5].
Figures I.1.2-3 (a) and (b) show the time evolution of the integrated power of the high frequency component (|f| > 200 kHz) of the O-mode reflectometer signal (P_refl.^H) and the electron temperature (T_e), respectively. The edge T_e at r/a ~ 0.85 was sharply dropped by the ablation of the pellet cloud at t = 6.32 s. The cold pulse induced by the pellet ablation propagated from the outside of the ITB into the ITB region. When the cold pulse propagated to the position of r/a = 0.54 in the ITB region at t = 6.325 s, P_refl.^H was drastically reduced with a fast time scale (< 5 ms), indicating the reduction of the density fluctuation. At this timing, T_e outside the ITB (r/a = 0.6-0.85) was decreased by the cold pulse propagation, while T_e in the ITB region (r/a < 0.54) was not changed due to the slower propagation speed of the cold pulse in the ITB region. Thus, the T_e gradient became large in the outer ITB portion (r/a = 0.54-0.64). The value of T_e at r/a = 0.54 largely decreased after the density fluctuation reduction, and the ITB structure seems to be destroyed in the outer ITB portion at t = 6.35 s as shown in Fig. I.1.2-3 (c).
The transient T_e response until 20 ms after the pellet injection (Δt = +20 ms) was simulated using 2 different models on χ_e. Here, Δt = 0 ms corresponds to t = 6.32 s. In the first case (power balance χ_e case), χ_e profile was assumed identical to the power balance χ_e at t = 6.3 s. In the second case (adjusted model χ_e case), the time evolution of χ_e was adjusted in order to reproduce the T_e time behavior by introducing a χ_e model depending on T_e and T_e gradient (∇T_e). The χ_e profile used in the power balance χ_e case is shown by the solid line in Fig. I.1.2-4 (a). Time behavior of the calculated T_e for the power balance χ_e case is shown in Fig. I.1.2-4 (b). In the outside of the ITB, the calculated T_e agreed well the measured one. The cold pulse did not propagate into the inner portion of ITB (r/a < 0.44) in the simulated time scale, which was consistent with the measurements. However, at r/a = 0.54 in the ITB region, the reduction of the calculated T_e was remarkably smaller than the measured one. The power balance χ_e cannot reproduce the measured T_e time evolution in the ITB region. In order to reproduce the time evolution of T_e, strong T_e dependence of χ_e was assumed as χ_e ∝ T_e^-2.3 in the outer ITB portion (r/a = 0.54-0.62). Weak negative T_e dependence of χ_e was also assumed as χ_e ∝ T_e^-0.5 outside the ITB (r/a > 0.63). Furthermore, in the inner ITB portion (r/a = 0.3-0.5), it was assumed that χ_e is inversely proportional to ∇T_e as χ_e ∝ ∇T_e^-1. The χ_e profiles at Δt = +10 ms and +20 ms after the pellet injection are shown by dashed and dotted lines, respectively, in Fig. I.1.2-4 (a). In this case, the value of χ_e decreased in the inner ITB portion and increased in the outer ITB portion depending on the T_e profile change. At Δt = +20 ms, χ_e decreased by a factor of 2 in the inner ITB portion and increased by a factor of 1.5 in the outer ITB portion. The time behavior of the calculated T_e well agreed with the measurement, as shown in Fig. I.1.2-4 (c). Figure I.1.2-4 (d) shows time evolution of χ_e together with the time evolution of P_refl.^H, which represents the ñ/n level in the range of k ~ 1 cm^-1. The value of P_refl.^H decreases by 80% of its total reduction during Δt = 0 ~ +20 ms within 5 ms, while χ_e decreases by 40% at r/a = 0.49 and only 2% at r/a = 0.46 (around the cut-off layer) within 5 ms. The time scale of the χ_e reduction (~ 15 ms) is larger than that of the ñ/n reduction (< 5 ms) and is similar to that the T_e profile change. The different time scales for the reductions of χ_e and ñ/n indicated the decoupling of the electron heat transport and the long-spatial-scale density fluctuations. In addition, the increase in χ_e for the outer ITB portion also supported the decoupling. This result was consistent with the power balance analysis result with fully recovered temperature profiles.

It is now widely recognized that a plasma rotation profile plays critical roles in determining the transport and the MHD stability in tokamak plasmas. In particular, for the operation of self-regulating systems, such as burning high beta plasmas, integrated plasma control schemes including the rotation profile from the edge pedestal to the core plasma regimes are required. Towards this goal, the physics basis on the driving mechanism of the plasma rotation and on the momentum transport has to be constructed [1.2-6]. The JT-60U plasmas heated with near-perpendicular neutral beam injection (perp-NBI) exhibit a ctr-rotation in the toroidal direction [1.2-7]. An inward electric field induced by a ripple loss of fast ions was considered as a candidate for the ctr-rotation in the peripheral region, because JT-60U had large toroidal field ripple. In order to reduce the toroidal field ripple, ferritic steel tiles (FSTs) have been installed inside the JT-60U vacuum vessel on the low field side [1.2-8]. Effects of the ripple loss of fast ions on the toroidal rotation velocity (V_t) have been investigated based on data with and without FSTs.
The increase in the co-rotation in the large volume (Vol.) configuration was observed by installing FSTs as shown in Fig. I.1.2-5 even with the similar temperature and density profiles. The ratio of ripple loss power to the input power on this discharge (I_p = 1.23 MA, B_T = 2.6 T, q₉₅ = 4.1, Vol. = 77 m³) reduced by installing FSTs from ~ 4% to ~ 1% for co-NBIs and from ~ 30% to ~ 10% for perp-NBIs, respectively. The magnitude of ctr-rotation increased with increasing the ripple loss power in the peripheral region. The ctr-rotation with perp-NBIs reduced by installing FSTs as a consequence of the reduction in the ripple losses. The driving mechanism of the ctr-rotation and the location of the driving source with the perp-NBI were investigated using the OFMC code and a beam perturbation experiment (I_p = 0.87 MA, B_T = 3.8 T, q₉₅ = 8.2, Vol. = 72 m³). In the perp-NBI modulation experiment, the large modulation amplitude and the small phase delay were observed in the peripheral region (0.7 < r/a < 0.9). From this result, the location of the driving source of the ctr-rotation was recognized in the peripheral region and this location agreed with the region where fast ion losses mainly take place in the OFMC code calculation.
The toroidal rotation profile in the core region was also discussed from the viewpoint of a moment transport by means of data from the beam perturbation experiment. The toroidal momentum diffusivity χ_φ and the convection velocity V_conv in the core region (0.2 < r/a < 0.65) were evaluated from the transient momentum transport analysis using the radial profiles of the amplitude of the modulated part of V_t (V_t0) and of the phase delay of V_t0. The toroidal momentum diffusivity χ_φ was comparable to the ion thermal diffusivity, and V_conv is about 5 m/s (inward direction) at r/a ~ 0.5 in this plasma as shown in Fig. I.1.2-6. In addition, the non-perturbative V_t profile in the core region can be almost reproduced by using χ_φ and V_conv evaluated from the perturbative V_t.

References
1.2-1 Sakamoto, Y., et al., Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, 2006) CD-ROM file EX/P1-10
1.2-2 Sakamoto, Y., et al., Nucl. Fusion 45, 574 (2005).
1.2-3 Suzuki, T., et al., Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, 2006) CD-ROM file EX/6-4
1.2-4 Takenaga, H., et al., Plasma Phys. Control. Fusion 48, A401 (2006).
1.2-5 Takenaga, H., et al., Plasma Phys. Control. Fusion 49, 525 (2007).
1.2-6 Yoshida, M., et al., Plasma Phys. Control. Fusion 48, 1673 (2006).
1.2-7 Koide, Y., et al., Plasma Phys. Control. Nucl. Fusion Research 1, 777 (1993).
1.2-8 Shinohara, K., et al., in Fusion Energy 2006 (Proc. 21st Int. Conf. Chengdu, 2006) (Vienna: IAEA) CD-ROM file FT/P5-32

In a fusion reactor such as ITER, the plasma operation with a high fraction of bootstrap current is expected to reduce the need for externally driven current. In such plasmas, however, neoclassical tearing modes (NTMs) would be excited. Since NTMs degrade plasma performance and sometimes cause a disruption, it is critically important to establish a scenario to control NTMs.

(1) Control of Growth of Neoclassical Tearing Mode by Central Co-ECCD
We investigated the possibility of the control of a 3/2 NTM by the electron cyclotron current drive (ECCD) at plasma central region in the same direction as the existing plasma current (co-ECCD) [1.3-1]. Since the co-ECCD decreases the safety factor in the central regime, the scale length of the current profile will decrease at the mode rational surface, which will act to stabilize the according to the modified Rutherford equation. In addition, since the co-ECCD will enhance sawtooth oscillations as previously demonstrated in JT-60U [1.3-2], some effect s are sawtooth oscillations, stabilizing or destabilizing the effect on an NTM, is expected.
In experiments, the behavior of a 3/2 NTM for no ECCD, 2-unit ECCD, and 4-unit ECCD was compared (Fig. I.1.3-1). Plasma parameters were as follows: plasma current I_p=1.5 MA, toroidal field B_t= 3.7 T, safety factor at 95% of flux surface q₉₅=3.9. Without ECCD, a 3/2 NTM appeared at β_N~1.6 (β_p=1.2) by the neutral beam injection with ~20 MW, and no sawteeth was observed throughout the discharge. For 2-unit ECCD (EC-driven current I_EC~70 kA), sawteeth appeared soon after ECCD, and a 3/2 NTM appeared at similar time as no ECCD case. While the growth rate of the 3/2 NTM for the 2-unit ECCD was twice as large as that for no-ECCD, the mode amplitude saturated at similar level. For 4-unit ECCD (I_EC~130 kA), the growth of a 3/2 NTM was suppressed and the mode amplitude was kept ~1/5 of the other cases. The mode amplitude was kept low even after the EC wave was turned off, and sustained for more than 10 times energy confinement time. In both 2-unit ECCD and 4-unit ECCD cases, the 3/2 NTM was not triggered by a sawtooth crash. Frequency spectrum of magnetic perturbations for the 4-unit ECCD showed that the mode frequency of the 3/2 NTM was modulated by a sawtooth crash while such a modulation was not observed for the 2-unit ECCD case. This suggests that a sawtooth crash could affect the 3/2 NTM. The value of β_N for the 4-unit was 6% higher than that for the 2-unit ECCD case, showing that it can be another candidate for the NTM control.

(2) Active Control of Neoclassical Tearing Modes toward Stationary High-Beta Plasmas
In previous JT-60U experiments, an m/n=3/2 NTM was completely stabilized by using a real-time NTM stabilization system, where the detection of the mode location and the optimization of the injection angle of the EC wave were performed in real time. Also, a preemptive NTM stabilization, where the EC wave was injected to the anticipated mode location before the mode onset, was shown to be effective. In addition, a high-beta plasma with the normalized β_N~3 was sustained for 5 times longer than the energy confinement time by suppressing NTMs.
Effects of the central co-ECCD on an m/n=3/2 NTM were investigated. Since the central co-ECCD modifies a current profile and also destabilizes sawtooth oscillations, the co-ECCD can affect the behavior of the 3/2 NTM. In experiments, the growth of a 3/2 NTM found to be suppressed by the central co-ECCD with the amount of ~10% of the plasma current. This result demonstrates the possibility of a coexistence of sawtooth oscillations and a small-amplitude 3/2 NTM.
Stabilization of an m/n=2/1 NTM by ECCD at the mode rational surface was also performed. In particular, effects of the ECCD location on the stabilization were investigated in detail. A complete stabilization was achieved when the misalignment of the ECCD location was less than about half of the full island width (~5 cm). Also the mode amplitude found to increase when the misalignment was comparable to the island width. After the finer tuning of the ECCD location, a 2/1 NTM has been completely stabilized only with one gyrotron power (~0.6 MW). The ratio of the EC-driven current density to the bootstrap current density at the mode rational surface, j_EC/j_BS, which is a measurement of the efficiency of the NTM stabilization, was as low as ~0.5.
Simulation of the stabilization of an m/n=2/1 NTM was performed using the TOPICS code combined with the modified Rutherford equation (Fig. I.1.3-2) [1.3-4].
The undermined coefficients in the modified Rutherford equation were determined by comparing with experimental results. The TOPICS simulations were found to well reproduce the stabilization and destabilization of a 2/1 NTM observed in JT-60U experiments with the same set of coefficients in the modified Rutherford equation. The TOPICS simulation also predicted that the ECCD width has a strong effect on NTM stabilization.

To realize an economical fusion reactor, the stabilization of the low-n kink-ballooning mode is necessary. The growth rate of the external kink-ballooning mode can be reduced with a close-fitting conducting wall, and the resulting mode, the so-called resistive-wall mode (RWM), has a growth time corresponding to the time constant of the relaxation of the wall current.

(1) Effects of plasma-wall separation on the stability of resistive wall modes
It is predicted that the RWM stability, in particular the growth rate, strongly depends on the plasma-wall separation. Therefore, to clarify the effect of the plasma-wall separation on the RWM stability, we have performed current-driven RWM experiments in the JT-60U ohmic plasmas [1.3-4]. In ohmic discharges, since a MHD stability is determined by only q-profile, the behavior of the MHD instability can be purely interpreted. In the experiment, the plasma was placed near the outer wall, and the plasma current was ramped up so as to destabilize the current-driven RWM. When the edge safety factor q_eff, which is q-value at about 97% of minor radius, was just below 3, a thermal collapse occurred due to an instability with a clear m/n = 3/1 mode structure. The growth time of this instability is about 10 ms which is the order of the skin time of the JT-60U wall τ_w = 10 ms. Therefore, the instability has been identified as a current-driven RWM.
To investigate the effects of the plasma-wall separation on the RWM, the plasma-wall separation was changed from shot to shot, and the clearance feedback control was utilized to keep the plasma-wall separation constant in a shot. In consequence, it is found that the growth rate has the strong dependence on the wall position as shown in Fig. I.1.3-3. Thus, the growth rate became smaller with decreasing the plasma-wall separation. Additionally, the observed growth rates have been compared with the AEOLUS-FT code, which is based on the resistive MHD equations with a resistive wall, and the model of the RWM dispersion relation. According to the AEOLUS-FT calculations, an m/n = 3/1 kink mode and an m/n = 2/1 tearing mode are unstable together under these experimental conditions. The dependence of the observed growth rate on the wall position is in qualitative agreement with an m/n = 3/1 kink branch, in contrast, it is quite different from an m/n = 2/1 tearing branch. However, in this comparison, the absolute value of the observed growth rates is about a factor 10 smaller than the dispersion relation with τ_w = 10 ms. Taking into account the time scale of the equilibrium evolution, the observed growth rate could be smaller than a liner growth rate [1.3-5]. Actually, the reevaluated growth rates are consistent with the RWM dispersion relation with τ_w = 10 ms. This point is required to investigate in future work.

(2) Identification of a Low Plasma-Rotation Threshold for Stabilization of the Resistive-Wall Mode
There are two different procedures for stabilizing RWM. The first is the feedback control stabilization using externally applied non-axisymmetric magnetic fields with coils in order to compensate for the diffusion of flux. The second is stabilization of RWM by the toroidal plasma rotation. Both theories and experiments imply that the critical rotation velocity is around 1%-2% of the Alfven velocity. The most significant problem in the previous experiment is that the investigation of the critical rotation is performed with the magnetic braking by adding an asymmetric magnetic field, the so-called error field. It has been pointed out that the error field plays a very important role in the destabilization of RWM and should be substantially reduced near the limit for the ideal mode without a wall, the so-called "no-wall limit". Since RWM stability itself is affected by the error field, which is the origin of magnetic braking, we should not use the magnetic braking in the investigation of critical rotation.
For investigating the critical rotation, we employ a weak reversed magnetic shear plasma with I_p = 0.9 MA and B_t = 1.58 T [1.3-6]. Internal inductance of this plasma is decreased to l_i = 0.8 to reduce the no-wall limit. The clearance between the plasma surface and the first wall on the outer midplane is d₀ = 20 cm, which is equivalent to a ratio of the radius of the first wall to the plasma minor radius of d/a = 1.2. A very low rotation threshold was obtained in the JT-60U in an investigation of the critical rotation for stabilizing RWM by controlling the toroidal plasma rotation with changing the combination of tangential neutral beams (NBs) without magnetic braking. The observed critical rotation is V_t ~20 km/s and corresponds to 0.3% of the Alfven velocity at the q=2 surface, much smaller than the previous prediction with the magnetic braking as shown in Fig. I.1.3-4. This low critical rotation does not increase as the beta value increases toward the ideal wall limit. Also an ITER relevant low rotation (~ 0.4% of the Alfven velocity) stabilization of RWM is demonstrated for 50 times the skin time of the first wall. These results indicate that for large plasmas such as in future fusion reactors with the low rotation, the requirement of the additional feedback control system for stabilizing RWM is much reduced.

Alpha particles play an important role in the plasma heating in burning plasmas. However, a high alpha particle pressure gradient can induce Magnetohydrodynamics (MHD) instabilities such as Alfv_n Eigenmodes (AEs) (toroidicity-induced AEs (TAEs)) or Energetic particle modes (EPMs). These MHD instabilities can cause the enhanced transport of alpha particles from the core region of the plasma, which could degrade the performance of burning plasmas. Moreover, lost alpha particles could also damage the first walls. Thus, the understanding of the alpha particle transport due to these instabilities is an important research issue for ITER. Then, several kinds of AEs and EPMs have been predicted theoretically and observed experimentally and their effects on energetic ions have been studied in several Tokamaks, ST and Helical devices. Recently, another type of AEs, in which the frequency is largely swept with the time scale of a few hundreds millisecond and then saturates as the minimum value of the safety factor (q_min) decreases, have been extensively studied. These frequency behavior can be explained by reversed-shear induced AEs (RSAEs) [1.3-7] and its transition to TAEs. Confinement degradation of energetic ions due to these AEs has been confirmed from the measurements of the total neutron rate as shown in Fig. I.1.3-5 [1.3-8]. In order to investigate the radial transport of energetic ions due to these modes, the line-integrated neutron emission profile are measured using a large neutron collimator array [1.3-9] and compared with the calculated value by a classical theory using a transport code TOPICS. As a result, it is found for the first time that energetic ions are transported from the core region of the plasma due to these AEs as shown in Fig. I.1.3-6. Further, changes in the energy distribution and the flux of the charge exchange (CX) neutral particle are measured [1.3-10] in order to investigate the energetic ion transport in the velocity space. The changes in the energy distribution of CX neutral fluxes suggest the radial transport is due to the resonance interaction between energetic ions and AEs. Thus, the results of the measurements of neutron emission profile and CX neutral particle flux indicate that the radial transport of energetic ions is induced by the resonance interaction between energetic ions and AEs.

References
1.3-1 Isayama, A., et al., Plasma Fusion Res. 1, 030 (2006).
1.3-2 Isayama, A., et al., Plasma Fusion Res. SERIES 5, 324 (2002).
1.3-3 Isayama, A., et al., Nucl. Fusion 47, 773 (2007)
1.3-4 Matsunaga, G., et al., Plasma Phys. Control. Fusion 49, 95 (2007)
1.3-5 Callen, J.D., et al., Phys. Plasmas 6, 2963(1999)
1.3-6 Takechi, M., et al., Phys. Rev. Lett. 98 (2007) 055002.
1.3-7 Takechi, M. et al., Physics of Plasamas 12, 082509 (2005)
1.3-8 Ishikawa, M., et al., Nucl. Fusion 47, 849 (2007)
1.3-9 Ishikawa, M., et al., Rev. Sci. Instrum. 73, 4237 (2002).
1.3-10 Ishikawa, M., et al., Rev. Sci. Instrum. 75, 3963 (2004).

Since the performance of ELMy H-mode plasmas can be determined by the performance of the pedestal plasmas, it is important to understand what physics determine the pedestal performance for the better prediction of pedestal plasmas in ITER. In addition to this, mitigation of the instantaneous ELM heat load should be established to achieve the required lifetime of divertor target plates in ITER. To address these issues, various experiments have been performed. Effects of toroidal field (TF) ripple on the pedestal performance and ELM characteristics were investigated by comparing with and without ferritic steel tiles (FSTs) using various plasma configurations including JET/JT-60U similarity configuration. As a co-toroidal rotation increases, the pedestal pressure tends to increase except for JET/JT-60U similarity configuration. Especially in a large plasma configuration where TF ripple without FSTs was large (~2%), an improvement of the pedestal pressure was observed even at the fixed toroidal rotation indicating distinct effects of the TF ripple. As for the size of type I ELMs, on the other hand, the normalized ELM size to the pedestal stored energy (ΔW_ELM/W_ped) decreased with increasing the counter rotation in all configurations. The fact that there was a systematic difference in the variation of ΔW_ELM/W_ped as a function of toroidal rotation indicated that the plasma configuration itself is one of important parameters to determine the ELM characteristics rather than the TF ripple [1.4-1]. The ELM frequency dependence in the grassy ELM regime (δ>0.5 and q₉₅>6) was investigated in terms of the toroidal rotation at the pedestal and total poloidal beta (β_p). As counter rotation increased, the ELM frequency clearly increased up to ~1400 Hz independent of β_p from 0.84 to 1.88. Even in the no-rotating plasma with balanced-NBIs similar to the ITER condition, a high ELM frequency of ~400 Hz has been observed without a large energy loss (less than 1% of the pedestal stored energy) [1.4-2]. QH-mode experiments as an inter-machine experiment with DIII-D was performed to understand the mechanism of the QH mode transition with co-NBIs. The new CXRS system revealed that similar depth of E_r well of ~40kV/m was observed both in the ELMy phase (just before an ELM) and the QH phase in JT-60U in contrast to DIII-D, where deeper E_r well of ~100kV/m has been observed. By making use of recent improvement of edge diagnostics with the fast time resolution, studies of ELM dynamics have been progressed. Just after a current ramp up, the higher pedestal pressure could be achieved together with larger type I ELMs, which may relate the improvement of the edge stability.

In the previous experiments, effects of the plasma rotation and the TF ripple on the ELM characteristics have been studied by changing the plasma configurations (δ_ripple~0.4% at small V_P~50 m³, δ_ripple~1% at middle V_P~65 m³, and δ_ripple~2% at large V_P~75 m³). Figure I.1.4-1 shows a relation of ΔW_ELM/W_ped with a V_T,ped under a P_SEP~5-6 MW at each plasma configuration, showing similar slopes in the dependence of ΔW_ELM/W_ped on V_T,ped. A reason for the narrower dynamic range toward the co-direction in the V_T,ped for a larger volume configuration is due to an offset toroidal rotation towards the counter-direction at the edge region due to larger losses of fast ions.
The most important point in Fig. I.1.4-1 is a systematic difference in the ΔW_ELM/W_ped (in other words, an offset of ΔW_ELM/W_ped) varying from ~9% (small-volume), ~6% (middle-volume), and ~4% (large-volume) at a given V_T,ped~-50 km/s, suggesting important roles of other parameters in determining the edge stability, such as TF-ripple, loss of fast ions, and plasma configuration. It should be noted that the pedestal pressure and/or the poloidal beta, β_p,ped, tends to decrease when the plasma volume increases. However, the decrease in ΔW_ELM is larger than that seen in W_ped, and hence ΔW_ELM/W_ped decreases when the plasma volume or δ_ripple increases.

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal (T_iped) on the ELM characteristics in JT-60U has been performed. The level of toroidal plasma rotation has been varied by using different combinations of tangential and perpendicular neutral beam injection (NBI). In the grassy ELM regime at high triangularity (δ) and high safety factor (q), the ELM frequency clearly increased up to ~1400 Hz, when counter (ctr) plasma rotation was increased as shown in Fig. I.1.4-3. The response of the ELM frequency was independent of poloidal beta (β_p) in the range 0.84<β_p<1.88 at δ>0.53. Even in non-rotating plasma with balanced NBIs, a high ELM frequency of ~400 Hz has been observed without large energy loss. When the frequency of the plasma rotation in the co-direction of the plasma current becomes higher than ~1 kHz, type I ELMs with a frequency of ~20 Hz have been observed. The achieved pedestal pressure and plasma confinement were similar both in plasmas with type I ELMs and in plasmas with grassy ELMs. The energy loss due to grassy ELMs was evaluated from the reduction in the electron temperature, and the ratio of the energy loss to the pedestal stored energy was less than 1%.

References
1.4-1 Kamiya, K., et al., Plasma and Fusion Research, 2, 005 (2007).
1.4-2 Oyaman, N., et al., Plasma Phys. Control. Fusion, 49, 249 (2007).

Studies of ELM plasma propagation in SOL have been progressed both at High- and Low-field-side [1.5-1]. Radial distributions of the ELM plasma was measured with three reciprocating Mach probes (synchronized sampling of 500kHz) in Type-I ELMy H-mode with frequency of 25-50 Hz. Multi (a few to several) peaks were generally observed in the ion saturation current (j_s) during ELM activities (_p). Radial propagation time (τ_⊥) from near-separatrix to the Mach probe location (r) was evaluated from time lag of the first large j_s peak after large _p start. ELM propagation to the midplane first wall (τ_⊥^mid = 50-90 μs at r^mid = 13cm) was faster than the parallel convection time to the outer divertor (τ_//^SOL-LFS ~100 μs), suggesting ELM plasma loading to the first wall. Radial distribution of τ_⊥^mid shows that the velocity (V_⊥^mid = r^mid/τ_⊥^mid) was increased from 0.4-1.5 km/s near separatrix to 1.5-3 km/s at the far SOL, as shown in Fig.I.1.5-1. Here, peak jsmid distribution was characterized by a wide e-folding length of 7.5 cm.
In the waveform of multi-peak j_s^mid, similar characteristics were seen for series of ELM events. Radial scale of ELM filaments was determined from duration of the large j_s^mid peak (δt_pk^mid), 4-10 μs. Corresponding radial scale (V_⊥^midδt_pk^mid) was 0.3-1.8 cm near separatrix (r^mid < 5 cm) and was increased to be 1 - 4 cm at the far SOL. Deposition image of particle flux to outer baffle and dynamics of the filament-like structure were also measured with new fast TV camera. Different time scales of ELM plasma transport were determined at the HFS SOL. Multi-peak j_s^mid was seen, however, short delay (< 30 μs) of large j_s^HFS was seen only in narrow region near separatrix (Δr^mid < 0.4cm). This would cause fast heat and particle transport to the HFS divertor simultaneously to that to the LFS divertor. On the other hand, in the far SOL at HFS, particle flux reached with larger time lag of the parallel convection time from the LFS SOL (τ_//^SOL-LFS =250-300 μs).
Fast measurements (20kHz sampling) of low ionization spectral lines such as three HeI, D_α and CII started for study of ELM plasma dynamics at the LFS divertor such as T_e, n_e, neutral and carbon ionization fluxes. Just after ELM event, T_e was increased, then it decreased due to large increase in recycled neutrals. At the same time, carbon yield became lowest.

Wall saturation has been investigated in 30s-ELMy H-mode discharges. Wall pumping rate was determined by a particle balance analysis and it decreased with a decay constant of several seconds and then became constant. After a few discharges with a density of 65% of the Greenwald density, the wall-pumping rate became negative, i.e. outgassing. The outgassing rate was correlated mostly with an increase in the tile surface temperature near the outer strike point. Thus, in high density discharges (80% of the Greenwald density), where heat loading to the strike point was largely reduced due to radiation loss, wall-pumping was observed over the long pulse [1.5-2]. The wall-pumping rate became larger than that explained only by the co-deposition of deuterium with carbon.
Divertor-pumping is effective to control the divertor plasma detachment and the main plasma density. Under the wall saturated condition (outgassing), it was found that the wall-pumping rate changed so as to maintain the main plasma density; outgas flux was decreased when the divertor-pumping flux was increased by opening the shutter of the divertor cryopumps. This fact suggests that the wall-pumping can be recovered to some extent (but still negative rate) in several seconds, i.e. dynamically retained changed by external control of the divertor pumping flux [1.5-3].

For high density divertor discharges, MARFE appears around the X-point, where the C IV brightness is largely peaked. Measurement of the absolute intensities of C IV lines emitted from this peak, using 2-dimential and wide-wavelength spectrometer system, could determine the population density ratio of the C³⁺ excited levels, and the result was analyzed by a collisional-radiative model [1.5-4]. The result of this analysis is shown in Fig I.1.5-2, indicating that the population densities of the excited levels were composed of the ionizing and the recombining plasma component with an electron temperature of 6.3 eV and an electron density of 7.8 x 10²⁰ m^-3. By using the collisional-radiative model with these evaluated parameters and the C IV brightness, the line-radiation power was evaluated: the line-radiation power from the ionizing and the recombining plasma component of C³⁺ was, respectively, 60% and 2% of the total radiation power (1.4 MWm^-2) measured by a bolometer along the chord at the X-point. It was also found that the volume recombination flux of C⁴⁺ is larger by two orders of magnitude than the ionization flux of C³⁺; the volume recombination is a main process to produce C³⁺, i.e. a main radiator in the X-point MARFE.
Two-dimensional structure of the volume recombination of deuterium, i.e. D⁺ and e^-, was investigated in the partially detached divertor. The deuterium Balmer-series lines ( D_α, D_β, ..., D_θ) were measured two-dimensionally with a spatial resolution of ~ 1 cm [1.5-5] and distributions of their emissivities were reconstructed with a tomography technique [1.5-6]. Emissivity ratio of D_β to D_α lines was compared to that calculated by the collisional-radiative model. This ratio could not be explained only by the excitation of the ground state deuterium by electron impact, indicating that the volume recombination contributed to the D_β emission. Fig. I.1.5-3 shows two dimensional distribution of D_β to D_α line ratio, and volume recombination process is dominant above the inner strike point with ~ 8 cm and ~ 4 cm, respectively, in the r- and the z-direction as indicated by the black curve. In this region, the electron density and temperature were evaluated to be ~ 5x10²⁰ m^-3 and < 0.3 eV, respectively, from the ratios of the D_α, D_β, ..., D_θ emissivities.

In the previous work [1.5-7], the chemical sputtering yield was evaluated by using 'Loss-Events/Photon' coefficients measured in PISCES-A. Because, however, the LEP coefficients depend on the transport loss of the hydrocarbons, which may depend on devices, the LEP coefficients, CH₄, CD₄, C₂H₄ and C₂H₆ loss events per CH, CD and C₂ photon, were recently measured carefully [1.5-8] in many tokamaks. By using these measured LEP coefficients, the chemical sputtering yield, reported in [1.5-6], was re-evaluated and is shown in Fig. I.1.5-4. Compared to the chemical sputtering yield in the previous report [1.5-6], the yield itself did not change. But the ion flux dependence of the yield, evaluated by the regression analysis with Y_chem Γ_ion^α, became weaker by 40%; α decreased from ~0.3 to ~0.2.

Movement and velocity of emission by dusts were measured with a fast visible TV camera (2-6 kHz) from tangential port [1.5-9]. In main SOL, many dusts with various directions were observed, in particularly, in the first NB shot after hard disruptions (large ΔW_dia) and overnight (~6 hours) GDC. In divertor, many dusts were produced at HFS, in particular, for high HFS strike-point case: large ELM heat and particle loading on thick deposition layers may enhance producing dusts. Fast velocity of toroidal or/and poloidal movement (0.2-0.5 km/s) was faster than that of radial movement (0.05-0.1km/s). Direction of toroidally moving dusts was mostly towards ion drift (I_p): the direction is consistent with the SOL flow measurement HFS and LFS. It was also found that direction of radial movement at LFS SOL above the divertor changed from inward to outward at ELM occurrence.

We have been developed a comprehensive SOL/divertor simulation code system for the interpretation and the prediction on behavior of SOL/divertor plasmas, neutrals and impurities [1.5-10]. The code system consists of (i) the 2D fluid code: SOLDOR, (ii) the neutral Monte-Carlo (MC) code: NEUT2D, and (iii) the impurity MC code: IMPMC. The integration code "SONIC" was almost completed and examined to simulate the SOL and divertor plasmas in JT-60U.
SOLDOR and NEUT2D codes have been applied to analysis of JT-60U experiments and divertor designing of JT-60SA. The X-point MARFE in high density experiments was examined. It was found that transport of carbon ions generated at the private dome are enhanced in the private plasma, causing the large radiation peaking near the X-point. The pumping capability of JT-60SA was evaluated through the simulation. A guideline to enhance the pumping efficiency was obtained in terms of the exhaust slot width and the strike point distance.
Code-code benchmark is one of very important collaboration of divertor/SOL topical group in ITPA (International Tokamak Physics Activity). Benchmark work between SONIC code (SOLDOR and NEUT2D) and SOLPS5.0, which has been used for the ITER divertor designing, started. We applied both codes on a typical L-mode divertor plasma on JT-60U, using each mesh structure and identical input parameters such as input power, plasma edge density, gas puff flux and divertor pumping speed. As a result, electron temperature and density in the SOL and divertor plasmas agreed fairly well.

Tungsten deposition in the divertor region eroded from the W-tiles installed in the outer divertor (P-8 section) was measured to study tungsten migration [1.5-11]. Poloidal distribution of tungsten surface density at the P-8 toroidal section is plotted in Fig.I.1.5-5, and compared to deposition profile of isotope carbon (¹³C) (a) and the normal carbon (¹²C) deposition profile (b). Here, carbon isotope was used for tracer of the carbon deposition just before machine ventilation in 2004. In the inner divertor tile, the deposition positions of tungsten and ¹³C are the same, indicating that both elements migrated near separatrix field lines and deposited. This was similar to the results in ASDEX-U [1.5-12]. For normal carbon deposition (sputtered from graphite tiles), carbon ions also migrated in SOL plasmas formed thick deposition layer in the upper side of the inner divertor, see Fig.I.1.5-5 (b).
On the outer wing of the dome tiles, deposition of ¹³C (puffed from the hole at the outer divertor tiles) and normal carbon deposition was increased near the exhaust slot similar to the case of tungsten deposition. The ratio of the tungsten to the carbon, however, was higher by an order of magnitude on the outer wing than that on the inner divertor. In addition, tungsten deposition on the outer wing at P-5 section is negligibly small compared with that at P-8 section (W-tile area). This suggests that the tungsten deposition on the outer wing tiles is highly localized near the W-tiles. The tungsten deposition could take place by inward drift motion following sputtering and ionization of tungsten atoms from the W-tiles or direct deposition of emitted tungsten atoms by disruption or ELMs etc. without ionization. Similar process could also occur for the carbon deposition.

Tracing isotope gas (¹³CH₄) was puffed into the outer divertor plasma at P-8 toroidal section SOL plasmas in 2004. After the ¹³CH₄ gas puffing experiment, carbon tiles in P-8 and P-5 (toroidally 60 apart from P-8) toroidal sections were extracted and the toroidal distribution of ¹³C deposition on the divertor tiles have been investigated [1.5-13, 1.5-14].
Poloidal distribution of ¹³C deposition density on the outer divertor tiles had a peak near the strike point. At the same time, ¹³C deposition at the outer strike point was peaked toroidally in the vicinity of ¹³CH₄ puffing nozzle with the density of ~10²¹ cm^-2. There was a large asymmetry between upstream and downstream sides along the field lines; heavy deposition in downstream side and little deposition in upstream side, suggesting that ¹³C ions were transported to the downstream by the plasma flow in the divertor region. Poloidal distribution of ¹³C deposition on the inner divertor tiles were investigated in both P-5 and P-8 toroidal sections, and the depositions were peaked near strike-point but at a private area. Determination of ¹³C transport through the private flux region (ion or hydrocarbon gas) will be necessary.
The poloidal distribution of ¹³C on both the inner and the outer dome tiles was similar to the carbon deposition distribution on the outer divertor: ¹³C influx on the lower side of the dome-wing tiles was larger than that on the dome top. ¹³C was also deposited on the poloidal side surfaces facing pumping slot on which ionized particles couldn't reach: at 3 mm depth from the plasma-facing surface. These results suggest transport of ¹³C neutral particles through the private flux region. Small amount of ¹³C was deposited on the toroidal side surfaces near the dome top, facing to the upstream side.

The total hydrogen isotope retentions depending on the tile location was investigated [1.5-15]. The hydrogen concentration was generally increased with the thickness of the co-deposited carbon layers, and was roughly classified into two different values. The lowest concentration of 0.02 in (H + D)/C was found in the co-deposited layers on the inner divertor tile, while the highest concentration of 0.13 in (H + D)/C was in the co-deposited layers on the bottom side of outer dome wing tile facing to the pumping slot. The hydrogen isotopes retention of the whole divertor area is roughly estimated to 4x10²⁴ atoms. Such low H+D retention in JT-60U compared to that in the other tokamaks is largely owing to higher temperature of the PFC tile surfaces.
Hydrogen retention in the codeposited layers on the collector probes under the divertor PFCs were summarized in Fig.I.1.5-6. The maximum D/H ratio of ~3.6 appeared underneath the dome region. The hydrogen (H+D) retention in the layers was ~1.4x10²³ atoms/m², and H+D/C was large, i.e. ~0.85 (average of two samples: ~0.75). On the other hand, no appreciable co-deposited layer was found on the collector probes installed underneath the outer baffle plate (No.5-10), and the hydrogen retention was almost background level. During the discharges, the temperature of the shadowed area was maintained at ~470 K (operation temperature), which was still low compared with the temperature of the surfaces of PFCs.

References
1.5-1 Asakura, N., et al. , Proc. 21st IAEA Conf., on Fusion Energy 2006 (Chengdu 2006) CD-ROM file EX9-2.
1.5-2 Nakano T., et al., Nucl. Fusion 46, 626 (2006).
1.5-3 Kubo H., et al., Proc. 21st IAEA Conf. on Fusion Energy 2006 (Chengdu 2006) CD-ROM file EX/P4-11.
1.5-4 Nakano T., et al., Proc. 21st IAEA Conf. on Fusion Energy 2006 (Chengdu 2006) CD-ROM file EX/P4-19.
1.5-5 Fujimoto K., et al., Trans. Fusion. Sci. Tech. 51, 247 (2007).
1.5-6 Fujimoto K., et al., Plasma Fusion Res. "Modification of Tomography Technique for Two-dimensional Spectroscopic Measurement in JT-60U Divertor Plasmas"(in press).
1.5-7 Nakano T., et al., Nucl. Fusion 42, 689 (2002).
1.5-8 Nakano, T., et al., 32nd EPS Plasma Physics Conference (Tarragona, 2005) CD-ROM file P-5.007.
1.5-9 Asakura, N., et al., Proc of workshop on Dusts in fusion plasma (Warsaw, 2007).
1.5-10 Kawashima, H., et al., Plasma Phys. Control. Fusion 49, S77-S85 (2007).
1.5-11 Ueda, Y., et al., J. Nucl. Mater. 363-365, 66 (2007).
1.5-12 Krieger, K., et al., J. Nucl. Mater. 241-243, 734 (1997).
1.5-13 Ishimoto, Y., et al., J. Nucl. Mater., "Transport studies of carbon impurities using 13CH4 gas puffing in JT-60U" to be published.
1.5-14 Nobuta, Y. et al., Proc. 34th EPS Plasma. Phys. Conf.,(Warsaw, 2007) CD-ROM file P024.
1.5-15 Hirohata, Y., et al., J. Nucl. Mater., 363-365 854 (2007).
1.5-16 Masaki, K., et al., Proc. 21st IAEA Conf., on Fusion Energy 2006 (Chengdu 2006) CD-ROM file EX/P4-14.

In order to extend the operation regime to high density, new pellet injector has been developed for longer duration and high frequency. In the previous pellet injector with a piston type extruder, the frequency and time of injection are limited by extrusion speed and pellet extruder. So, in order to produce pellets continuously, the pellet injector using the screw type extruder has been developed. After the extruder is modified from piston type to screw type, the production of ice rod has been tested. The continuous pellet extrusion up to 330 sec was achieved using the screw extruder. Figure I.2.1-1 shows the photograph of continuous production of pellet. In future, after the design for vacuum and heat insulation the necessary modification for production of good quality ice rod and test will be done.

A purpose of the secondary cooling system is removal of the heat in the systems of torus, power supply, heating device etc by circulating water. A floating body is contaminated in the circulating water. In order to remove a floating body in cooling water, a sand filer is used. The floating body attached to the sand in the bath. The sand is washed by much water and medicine periodically (three times per week in JT-60 operation period). The expensive medicine is needed to keep the quality of the water. Since running cost using sand filter is expensive, the new filter method is introduced. The filter is washed automatically in the new method. The schematic diagram of the new method is shown in Fig. I.2.1-2. The filter consists of a screen mesh which is no replacement. When the clog is happened and difference pressure become bigger than the threshold value, the filter is washed. Amount of needed water in the case of screen mesh is one sixth of the water compared with that in the case of sand filter. Moreover, the expensive medicine is not needed; it is expected to reduce the running cost.

The carbon tiles were installed in order to guard the vacuum vessel. The carbon tiles in the divertor outer dome were dropped on 16th June in 2006. The cause of dropout is that heat from plasma flowed the tiles in divertor outer dome. The heat was concentrated to the bolt and nut that is concentrated between the tile and the basement of tile. The nut was melted by the heat flow and the tile was dropped.
The heat per shot was estimated to be 0.12~0.3MJ and the temperature increment is estimated to be 150C -370C by taking account of heat ratio between in and out divertor and heat ratio between divertor and dome. Since the temperature can be estimated to be increased to be 1400C by continuous 3-7 shots, the ferritic steel tiles may be melted.
We take measures against the melting bolt. In order to escape the heat, the carbon sheets will be inserted between tile and the basement that will be made newly. The carbon sheets also give the good heat contact between basement and tile.
The looseness of bolt and melting were checked in all ferritic steel tiles (total 1122 tiles). The looseness of bolt or melting was found in 45 the ferritic steel tiles. Especially many ferritic steel tiles near the horizontal port were found.

In the plasma operation, an abnormal plasma shape control occurred during plasma discharge due to abruptly drift error of integral result of magnetic sensors, especially rogowskii coil for plasma current calculation. We have carried out the repetition test to check the drift error phenomenon in integrator. From the impulse noise test result using the noise simulator, the cause of the drift error of integrator is considered the noise from AC power supply line or grounding.
Plasma current is precisely calculated by the line integral of magnetic field along closed line enclosing a plasma on the CCS (Cauchy-Condition Surface) computer system. For the realization of precise plasma current measurement, we have developed a new plasma current calculation formula using the CCS method [2.2-1] in real-time feedback control. High accuracy plasma shape and position control have been achieved using a plasma current of the CCS method in the plasma operation. Therefore, it is possible to measure a plasma current without measurement of rogowskii coil.

ITER remote experiment is planned to be tested by using JT-60SA at Japanese remote experiment center (REC, Rokkasho-mura, Aomori-ken, Japan) as a part of the ITER-BA project. Functions required for JT-60SA remote experiment are monitoring of discharge sequence status, handling of discharge parameter, checking of experiment data, and monitoring of plant data which is same as the existing JT-60 Human Interfacing System (HIFS). The HIFS is now exclusive to the on-site user in NAKA site due to network safety.
To realize remote experiment of JT-60SA, some functions should be executed from the REC via the Internet. However, the existing JT-60 HIFS functions have been designed only for on-site use for network security reasons. Therefore, we have developed the Java-based HIFS using Java language for remote experiments. Fig. I.2.2-1 shows the HIFS overview for JT-60SA remote experiments. In this system configuration, newly introduced remote experiment server (RES) is connected to the existing JT-60 site-wide LAN, and supervises both of Java application and server application software. A Java application program which is stored in RES is automatically downloaded and executed in users PC when it is selected by user through RES web home page. In this case, a Java application is called as a "Java applet".
JT-60 on-site users can use the HIFS for remote experiment by direct access to the web home page of RES from the web browser of individual users PC. The same HIFS environments are provided for the remote site users by connecting to the ITBL server via the Internet. Authentication program installation approval by JAEA is required for the remote site users to connect ITBL (Internet Technology Based Laboratory) server. The combination of Java-based software and ITBL has improved remote site experimental environment to the same level as JT-60 on-site.
New Java-based HIFS was successfully demonstrated in JT-60 remote experiments from Kyoto University in 2006.

For safety and efficient operation of a fusion device, appropriate information should be provided to the personnels participating in the experiment. As the first step along this requirement, we have newly developed highlight screens showing the plasma movie, the status of discharge sequence, the discharge schedule, the present status of trouble, etc., on the large (65 inch) monitor TV in front of the central consoles. These screens are automatically switched over according to the timing signal of JT-60 discharge sequence and/or the operator's request. This system is composed of two systems: The new real-time plasma visualization system (RVS) [2.2-2] and the discharge information management system (DIMS). We call this system the supervisory discharge operation monitoring (SDOM) system.
Hardware configuration of SDOM is illustrated in Fig. I.2.2-2. While executing the discharge experiment, the discharge information (highlight screen) which was provided to the personnel related to experiment was outputted by SDOM with the large monitor TV in front of the central consoles. These highlight screens can be seen with the television at the room in addition to the JT-60 central control room.
The highlight screens displayed by SDOM have been supported four: (1) the latest discharge parameter and schedule, (2) the JT-60 manager's information, (3) the status of discharge sequence, and (4) the real-time plasma movie. DIMS collects the necessary information from the JT-60 control system, and automatically changes into highlight screens according to the timing signal of JT-60 discharge sequence and/or the session leader's request as shown in Fig. I.2.2-3. This system was developed using the latest JAVA language with only one personal computer. Each element is briefly explained below.
1) The latest discharge parameter and schedule
   SDOM shows the highlight screen which is the latest discharge parameter and schedule.
2) The manager's information
   If SDOM receives the JT-60 session leader's request, this system outputs the characters which are inputted at the center of screen.
3) The status of discharge sequence
   When the "discharge sequence start" signal is received, this system change into the highlight screen which is the status of discharge sequence. This screen is shown until this system receives the "One minute before discharge" signal.
4) The real-time plasma movie
   When the switcher receives the "One minute before discharge" signal, it changes into the highlight screen which is the real-time plasma movie. The plasma movie is composed of video camera picture looking at a plasma, plasma shape computer graphic (CG), and magnetic probe signal as a sound channel. A plasma shape CG is calculated in every 1 ms using the shape reconstruction method (the CCS (Cauchy-Condition Surface) method) by the plasma shape reconstruction system.
For the development of the SDOM system, JT-60 is able to execute the experiment on safety and efficient operation, and the physicist can do the latest discharge analysis quickly. Since this system was developed using the JAVA language, it can easily add the function according to the various requests in the future.

References
2.2-1 Kurihara, K., Fusion. Eng. Des. 51-52, 1049 (2000).
2.2-2 Sueoka, M., et al., "The Plasma Movie Database System for JT-60," Fusion Eng. Des., (2007) to be published.

Annual inspections and regular maintenances for the power supply systems have been conducted to maintain availability of high power operations as shown in Table I.2.3-1. In addition, the renewals and maintenances for the power supply systems have also been conducted as measures for the superannuated power supply systems. These renewals and maintenances are shown in Table I.2.3-2. These activities contributed to achieve safe operation of the power supplies.

However, a few troubles of the power supply systems have happened in JT-60 plasma experiment. As an example, trouble of the phase controller for the thyristor converter of the PFPS is mentioned. In plasma discharge operation on 21^st, August, 2006, current unbalance (61D7V) among thyristor converters PSV22A of vertical field coil power supply occurred in initial coil current energizing period. Figure I.2.3-1 shows the circuit configuration of PSV2. 61D7V is to detect current unbalance of 1.09kA in 100ms between 25% of current value measured by DCCT2V for PSV2 total current and measured value by DCCT for PSV22A current. As the reason of the current unbalance, it was found that a print board of digital phase controller to control the timing to trigger thyristor gate of PSV21C had fault in its output pulse for triggering. As the fault caused that PSV21C was operated to output reverse voltage, which was negative voltage to cancel out by chance the positive voltage output by normally operated PSV22A, any current did not flow through PSV21C and PSV22A. The print board in fault was exchanged for a spare print board with checking normal output pulse for triggering. It was found that the data processing ROM on the print board in fault, which is to output pulse for triggering, was damaged by aging deterioration in use.

In coil energizing operation on 17^th, April, 2006, thyristor element failure of thyristor converter PSF1 occurred. The reason of the failure was that the rapid fuse, which is to protect the corresponding thyristor element against over current etc, was blown out. The ratings of rapid fuse are 2100V, 800A and I²t=4×10⁶A²s, and the circuit configuration of PSF is shown in Fig. I.2.3-2.
The current waveform with the PSF1 thyristor element failure is shown in Fig. I.2.3-3. The failure occurred just before current-zero of PSF1 with terminating circulating current control. And also, the enlarged current waveform shows that the current rose again up to 32kA in 90ms and that the rising current included swinging part which had the frequency equal to that of primary side AC voltage of thyristor converter PSF1. As the result, it would be possible that the corresponding thyristor element kept conducting condition whereas PSF1 terminated circulating current control. Therefore, PSF2 was short-circuited equivalently and large current flowed through PSF1 resulting in rapid fuse being blown out. I²t value of one rapid fuse of PSF1 in the failure is estimated as 7.5×10⁶A²s, which exceeds the rated value of 4×10⁶A²s. Thyristor element kept conducting condition because circulating current control was terminated without sufficient reverse voltage to turn off thyristor. To supply sufficient reverse voltage on thyristor in circulating current control, the current level to terminate circulating current control of PSF has been lowered.

The negative-ion-based NBI unit was originally designed to inject 10 MW D⁰ beams for 10 s using two large negative ion sources, each of which is designed to produce 22 A, 500 keV D^- ion beams. In the campaign of 2006, the beam pulse length is strongly required to be extended up to 30 s in order to study quasi-steady state plasma in JT-60U. The estimation of the power supply capacity indicated that the injected beam power was limited to be < 4 MW for the injection time of 30 s. Under this limitation, a long pulse injection of the neutral beams was carried out by clarifying operational ranges for a stable voltage holding capability and an allowable grid power loading in the JT-60 negative ion source.
To choose the stable acceleration voltage for the long pulse injection, a correlation between the voltage holding capability and a light intensity of cathode luminescence from the Fiber Reinforced Plastic insulator was carefully examined. As the result, the stable voltage holding capability without the beam acceleration was < 420 kV where the light was sufficiently suppressed [2.4-1]. Since the stable acceleration voltage with beam acceleration was 70-80% of that without the beam acceleration, 320kV-340 kV was chosen as the beam acceleration voltage (V_acc).
To attain the beam pulse length of > 30 s, the grid power loading should be reduced to < 1 MW where the surface temperature of the grid is calculated to be lower than the re-crystallization temperature (250 C) of the grid material (oxygen-free cooper). The extraction voltage (V_ext) and the arc power (P_arc) were tuned to minimize the grid power loading while the injection power was maximized. At V_acc=320 kV, V_ext=5.5 kV and P_arc=180 kW, the maximum grid power loading was on the grounded grid, and was allowable level of 650 kW. The D^- ion beams of 30 A were produced with two ion sources and neutralized to 3.2 MW D⁰ beams by a gas neutralizer.
The pulse length of 3.2 MW D⁰ beam was extended step by step, and finally reached up to 21 s as shown in Fig.I.2.4-1. This is the first long pulse injection of > 20 s in a power range of > 3 MW [2.4-2]. The achieved pulse length was limited by the surface temperature of the beam scraper made of molybdenum without water cooling. The beam pulse length is to be further extended by increasing the interlock level.

In addition to N-NBI unit, 11 positive-ion-based NBI units, each of which has the capability of 2 MW, 10-s D⁰ injections, were modified to achieve the 30-s injection. Four tangential units were upgraded by mainly increasing the capacity of the power supplies so as to extend the injection time from 10 s to 30 s while the beam injection power was kept to be 2 MW. Using these tangential units, 30-s injection was demonstrated at a total injection power of 8 MW. For 7 perpendicular units, injection timing of each unit were optimized to maximize the injection power for the 30-s pulse. In 2006, by using the modified tangential units and the perpendicular units, ~ 800 shots were reliably injected to meet the requirements of plasma physics. This significantly contributed to the study on quasi-steady state plasmas in JT-60U.

The design studies of the power supplies (1), beamline (2) and magnetic shielding against the stray field (3) for JT-60 SA were progressed.

(1) Power supply
For the power supplies in the P-NBI system, the protective characteristics and the thermal capacity of the power supplies were examined. From the curve of the protective characteristic, the allowable acceleration voltage and current are < 86kV and < 55 A for the 100-s injection, respectively. At these acceleration voltage and current, the injected beam power for one unit is 2 MW, which satisfies the design value of JT-60 SA. The design study on the thermal capacity of the power supply shows that the water cooling breeder resistance of the acceleration power should be upgraded to achieve the 100-s injection at 2 MW.
For the power supplies in the N-NBI system, the inverter capacity for switching of the acceleration power supply is required to be upgraded for 100-s injection. Since the present GTO thyristors are not available due to stop of the manufacturing, a combination of present GTO and a new element is newly designed for 100-s operation. The IEGT (injection enhanced gate transistor), whose performance is reaching in the GTO handling power with lower loss, is a candidate for the new element. The R&D of the combination of GTO and IEGT will be done before the construction of the upgraded N-NBI power supply system. Other power supplies are also modified to upgrade their thermal capacities by attaching additional cooling fans and exchanging the circuit breakers for large ones.

(2) Beamline
The modification of the NBI beamline for JT-60SA was studied by a 3-D Computer Aided Design (CAD). The disassembly of the existing NBI system, the possibility for a 0.6 m downward shift of N-NBI beamline, and possibility for the connection of the existing P-NBI unit with cryostat in JT-60SA were urgently examined in this year.
For the disassembly, while the number of the removed components was minimized, the cost-effective disassembly procedure was proposed. The 0.6 m downward shift of the N-NBI unit can be realized only by newly manufacturing the shorter support structures for the N-NBI beamline. For the connection of the exiting P-NBI unit with cryostat in JT-60SA, the existing beamline of the P-NBI unit can be connected by newly manufacturing shorter drift duct without significant modification of the beamline

(3) Magnetic shielding against stray field
In JT-60SA, the maximum plasma current is designed to be 5.5MA, where the stray magnetic field from the reactor at the exit of the neutralizer cell on the perpendicular P-NBI unit is about 0.1 T. This is about three times larger than that in existing JT-60U. The present NBI system has canceling coils and a double-layer passive shield that is composed of mild steel as an outer layer and μ-metal as an inner one. To reduce the stray field less than ~0.03 mT on the perpendicular P-NBI unit, a canceling coil is additionally designed. It is also designed that the thickness of the outer passive shield is also increased from 2.6 cm to 10 cm. The beam simulation code shows that the beam deflection is about 3 mrad at the exit of the neutralizer cell, which allows the beam transmission to be the same level as that in the present P-NBI unit. The magnetic field at the plasma center, disturbed by magnetic shields of one perpendicular P-NBI unit, is estimated to be about 0.06% of that without the magnetic shields of P-NBI unit.

References
2.4-1 Hanada, M., et al., IEEE Transactions on Dielectrics and Electrical Insulation, 14, 3, 2007
2.4-2 Hanada, M., et al., Proc. of 21^st Int. Conf. on Fusion Energy 2006, IT/2-3Ra ( Chengdu , 2006 ).
2.4-3 Usui, K., et al., "Study on modification of power supply system for long pulse operation on JT-60 positive ion-based NBI", JAEA-Technology 2007-024, 2007 (in Japanese)
2.4-4 Mogaki, K., et al., "Preliminary design of beamline components for JT-60SA NBI heating system", JAEA-Technology 2007-025, 2007 (in Japanese)
2.4-5 Ikeda Y., et al., Fus. Eng. Des., to be published.

Performance of the JT-60U radio-frequency (RF) heating system has been constantly improved to extend the parameter region of experiments such as sustainment of high performance plasmas for a few tens seconds. In FY 2006, major improvements of the JT-60U RF heating system were extension of the pulse duration and high frequency power modulation of the electron cyclotron heating (ECH) system and raising the performance of the lower hybrid (LH) system.

Pulse duration of the ECH system is being tried to extend to enhance the plasma performance in the recent experiment campaign in JT-60U focusing on long sustainment of high performance plasmas. Improvements of the gyrotron and developments of advanced operation techniques are keys to extend the ECH pulse. A difficulty to the pulse extension was to keep the oscillation condition against decreasing collector (electron beam) current because of cathode cooling by continuous electron emission. The techniques of controlling heater current and anode voltage during the pulse developed by FY2004 [2.5-1] were refined [2.5-2] and pulse duration of 21 s at 0.4 MW (at gyrotron) has been achieved.

The JT-60U ECH system has a unique feature to realize the power modulation by controlling the anode voltage of the triode gyrotron without chopping the main acceleration voltage. Power modulation operation at relatively low frequencies below 50 Hz was practically used in order to measure heat conductivity of the plasma to investigate plasma confinement. The typical depth of the modulation was 80% at the modulation frequency range of 12.2 Hz to 500 Hz. However in JT-60SA or ITER, higher modulation frequency of ~5 kHz will be required to stabilize neoclassical tearing mode (NTM). As a result of a fine optimization of the anode voltage and impedance of the high voltage circuit, the modulation frequency of 3.5 kHz with the modulation depth of 84 % has been achieved so far in short pulse < 0.06 s. The modulation frequency up to 3 kHz is available in the pulse widths of the practical operation ~ 0.4 s as shown in Fig. I.2.5-2. As a next step, replacement of the parts in the anode voltage divider circuit is planned to achieve higher modulation frequency.

LH experiments such as real time control of plasma current profile were performed using LH system, especially with 8-carbon grills around LH launcher mouth since August 2003. After one-year operation, no major damage was observed on the carbon grills itself, on the other hand some base plates as an attachment for the grill were melted. Main cause was malfunction of the arc detector to protect LH antenna mouth at RF breakdowns. Therefore for suppression of damages around LH antenna mouth, checking lump of the arc detector is improved and detecting speed is increased. And a new protection device using CCD-camera is also introduced. The 6-modules without carbon grill are used for LH experiments, stable injections are obtained as 1.1 MW for 20 s (18.64 MJ) in 2006.
On the other hand, to improve insufficient electrical contact along the carbon grill, a new structural carbon grill is developed at a test bench by using "a diffusion bonding method". This new carbon grill is combined with a pedestal made of stainless steel. RF connection point exists between the pedestal and LH launcher mouth. In the RF connection point, a hollow tube type RF contactor has been used instead of a fine spring RF connector, the test module for the new carbon grill shows enough power capability of 0.5 MW - 10 sec at the test bench as shown in Fig. I.2.5-3. This technology will be applicable to the LH launcher in JT-60U.

References
2.5-1 Moriyama, S., et al., Fus. Eng. Des., 74, 343 (2005).
2.5-2 Fujii, T., et al., J.Phys.: Conference Series, 25, 45 (2005) .

Profiles of the ion temperature and the plasma rotation have been measured with the significantly high spatial resolution using the new charge exchange spectroscopy (CXRS) system with the space modulation optics, as shown in Fig. I.2.6-1. In this modulation CXRS system, the object lens (f = 20mm) in front of the fiber array is mounted on the Piezo nano-translation stage which has a travel range of 0.5 mm. The observation points are scanned by ± 31mm (corresponding to ± 24mm in radius). The light of the charge exchange line from the other end of the 31ch optical fiber array is led to the entrance of the Czerny-Turner spectrometer with an f = 400mm F2.8 camera lens and a 2160/mm grating. The back illuminated electron multiplying 512 x 512 pixel (16 x 16 μm) CCD camera is used as a detector to achieve a high frame rate, f_ccd, up to 400Hz with 16 pixel vertical binning.
The radial points of the measurement can be multiplied by M [=f_ccd /(2 f_m) ] if the plasma is in the steady state, where f_ccd is frame rate of CCD detector of the spectrometer and f_m is a modulation frequency of Piezo stage. By choosing f_ccd = 200Hz and f_m = 10Hz, the radial profiles of the ion temperature and the plasma rotation with 310 points are obtained every 50ms. In order to derive the first and second derivative of the ion temperature from the modulation component of the time evolution of the ion temperature, Fourier analysis technique has been developed [2.6-1]. This technique will provide significant impact in the transport analysis, because the temperature gradient is directly derived. Moreover the radial profiles of the second derivative (curvature) in the region of internal the transport barrier (ITB) are measured with the modulation charge exchange spectroscopy [2.6-2].

In JT-60U divertor plasmas, the deuterium Balmer-series line emission has been measured with a wide-spectral-band spectrometer, which has 92 viewing chords (vertically 60 chords, horizontally 32 chords) with a ~ 1 cm spatial resolution and covers a spectral range of 350 - 800 nm. Two-dimensional spatial distribution of the Balmer line intensities has been reconstructed [2.6-3] using a computer tomography technique (maximum entropy method). It is very important to reduce the false geometric pattern produced in the reconstructed profile as small as possible.
In order to compare the false geometric pattern, (a) the square grid and the experimental view lines (S-E combination), and (b) the parallel grid and the experimental view lines (P-E combination) were applied for the reconstruction of the test profiles. The better results were obtained with the P-E combination than those with the S-E combination. The peak value in the reconstructed profile was 65 % of the test and the normalized error sum of squares, ΔE was 0.50 with the P-E combination, while the peak value was 44 % and ΔE is 0.71 with the S-E combination [2.6-4].
The two dimensional Dγ emission profile, was obtained with the P-E combination for the detached plasma in the inner divertor, as shown in Fig. I.2.6-2.

A high-throughput polarization interferometer is being developed to demonstrate the utility of Fourier transform spectroscopy for Thomson scattering diagnostics. Fourier transform spectroscopy is a measurement technique based on measurements of the temporal coherence of a light source. In general, the main advantages of the Fourier spectroscopy are: (1) high throughput optics (i.e. interferometer), (2) improvement of S/N ratio because the interferometer observes the whole wavelength area at the same time. More importantly, for this application, the method can be implemented in a simple and compact.
A prototype of polarization interferometer as the Fourier transform spectrometer was developed as shown in Fig.I.2.6-3. Target T_e and n_e ranges of the prototype polarization interferometer are < 1 keV and > 5x10¹⁸ m^-3, respectively. To compare between the numerical simulation and the experiment, an initial test using a blackbody radiation source was carried out. When the temperature of the blackbody radiation source was changed, the fringe contrast was measured. As a result, the magnitude of the change in fringe contrast agrees with the numerical calculation [2.6-5]. Proof-of-principle tests was carried out in TPE-RX reversed-field pinch (RFP) machine. The electron temperature was successfully measured by the polarization interferometer in TPE-RX PPCD discharges. The temperature of the polarization interferometer nearly agrees with that of the polychromator. The validity of this method was proved for the first time.
This work is supported in part by Grant-in-Aid for Scientific Researches on Priority Area "Advanced diagnostics for burning plasma" from MEXT (No. 18035016).

The design target values of the ion gun for Zeeman polarimetry on JT-60U, which will be installed in 2007, are the beam energy of 30 keV, the beam current of 10 mA and the beam divergence angle within 0.13 degrees. The ion gun consists of the ion source, the acceleration electrodes, the electron suppresser, the Einzel lens, the XY deflectors and the neutralizer. Performance of the prototype ion gun has been investigated by using the TriComp beam simulation code (Field Precision), which takes into account of the space charge effects, as shown in Fig. I.2.6-4. Three different initial beam profiles at the ion source are assumed in the simulation, which are peaked, flat and hollow profiles. Each profile is feasible due to the non-uniform beam emission at the ion source, which is caused by the ion source heating characteristics, the extraction field profile, and the coating quality of the b-eucryptite.
The simulation results show the beam loss of 50% caused by the clash to the electrode such as the cathode and the neutralizer. As a whole, the peaked profile has better performance in terms of the beam focusing and efficiency. The beam current about 0.9 mA is obtained at the observation area when the extracted current is 15 mA [2.6-6].

References
2.6-1 Ida, K., et al., "Measurement of derivative of ion temperature using high spatial resolution charge exchange spectroscopy with space modulation optics," submitted to Rev. Sci. Instrum.
2.6-2 Ida, K., et al., "Transition between internal transport barriers with symmetric and asymmetric curvature of temperature profiles in JT-60U tokamak plasmas," submitted to Phys. Rev. Lett.
2.6-3 Fujimoto, K., et al., Transactions of Fusion Science and Technology 51, 249 (2007).
2.6-4 Fujimoto, K., et al., "Modification of Tomography Technique for Two-dimensional Spectroscopic measurement in JT-60U Divertor Plasmas", to be published in Plasma Fusion Res. (2007).
2.6-5 Hatae, T., et al., "Development of Polarization Interferometer Based on Fourier Transform Spectroscopy for Thomson Scattering Diagnostics," to be published in Plasma Fusion Res. (2007).
2.6-6 Kojima, A., et al., "Numerical Simulation of a High-Brightness Lithium Ion Gun for a Zeeman Polarimetry on JT-60U, " to be published in Plasma Fusion Res. (2007).

The JT-60SA project is a combined project of JA-EU Satellite Tokamak Programme under the Broader Approach (BA) Programme and JAEA's programme for national use, with equal exploitation opportunity between the two programmes [3.1-1].

The mission of the JT-60SA project is to contribute to the early realization of fusion energy by its exploitation to support the exploitation of ITER and research towards DEMO, by addressing key physics issues for ITER and DEMO.

A bird's eye view of JT-60SA is shown in Fig. I.3.1-1. Typical parameters of JT-60SA are shown in Table I.3.1-1. The maximum plasma current is 5.5 MA with a relatively low aspect ratio plasma (Rp=3.06 m, A=2.65, κ₉₅=1.76, δ₉₅=0.45) and 3.5 MA for an ITER-shaped plasma (Rp=3.15 m, A=3.1, κ₉₅=1.69, δ₉₅=0.36). Inductive operation with 100s flat top duration will be possible within the total available flux swing of 40 Wb. The heating and current drive system will provide 34 MW of neutral beam injection and 7 MW of ECRF. The divertor target is designed to be water-cooled in order to handle heat fluxes up to15 MW/m² for long time durations. An annual neutron budget of 4x10²¹ neutrons is foreseen and will be subject to the agreement by the responsible legal authority and local government [3.1-1].

The main features of JT-60SA device are as follows;
- JT-60SA is a fully superconducting tokamak capable of confining break-even equivalent class high-temperature deuterium plasmas.
- JT-60SA allows the exploration of configuration optimization for ITER and DEMO with a wide range of plasma shapes (elongations and triangularities, shape factor S=q₉₅I_p/(aB_t)) and aspect ratios (A=R/a down to ~ 2.6) including that of ITER, with the capability to operate in both single and double null configurations.
- JT-60SA allows the exploration of ITER relevant high density plasma regimes well above the H-mode power threshold with 41 MW high power heating, including dominant electron heating by 110&140 GHz ECRF and 500 keV N-NBI. - JT-60SA allows the study of power and particle handling at 41 MW for 100 s with top and bottom water-cooled divertors compatible with maximum heat flux of 15 MW/m².
- JT-60SA allows the exploration of full non-inductive steady-state operation with 10 MW/500 keV tangential NBCD and 7 MW of ECCD.
- JT-60SA allows exploitation of high beta regime with stabilizing shell covered with ferritic plates and internal RWM stabilizing coils and high power H&CD system.
- JT-60SA is equipped with remote handing system to allow maintenance of in-vessel components compatibly with the planned annual neutron yield of 4x10²¹ neutrons/year.

The duration of the Satellite Tokamak Programme will be 10 years, including 3 years for commissioning and operation. The time critical components are the cryostat base and toroidal field coils. Major milestones are start of tokamak assembly in 2011, completion of tokamak assembly early in 2014, and first plasma in 2014.

References
3.1-1 Kikuchi, M., et al., Proc. of the 21^st IAEA Fusion Energy Conference (Chengdu, 2006) CD-ROM file FT/2-5.

Optimization of operational scenarios, such as full current drive (CD) ones, assessment and optimization of MHD controls, for the resistive wall mode (RWM) and the neo-classical tearing mode (NTM), and evaluation of heat/particle handling have progressed [3.1-1, 3.2-1].

Optimization of full CD plasmas has progressed [3.1-1, 3.2-1]. A full CD plasma with I_p = 3 MA (B_t = 2.27 T, q₉₅ = 5.6, β_N = 3.6, β = 4.2%,. f_BS = 0.55) is expected at /n_GW of 0.58 for H_H98y2 = 1.33 with total heating power of 41 MW. At this toroidal field, both 110 and 140 GHz ECRF can be absorbed at the second harmonics resonance. Adjusting the injection angles (toroidal and poloidal angles), both 110 and 140 GHz ECRF waves are absorbed at r/a~0.55 and generate EC-driven current of 0.14 MA. A full CD operation with a higher normalized density of /n_GW = 0.86 (= 5.0x10¹⁹ m^-3) is expected at I_p = 2.4 MA for H_H98y2 =1.33 (B_t = 1.82 T, q₉₅ = 5.7, β_N = 4.3, β = 4.9%) with 41 MW injection power as well. In this case, the 140 GHz ECRF is absorbed at the third harmonics then does not drive significant current, while the 110 GHz wave is absorbed at the second harmonics resonance. The fraction of the bootstrap current reaches 0.69. The plasma equilibrium, the current profiles (BS, EC, BD, OH and total) and Te Ti ne and q profiles of this high density high β_N full-CD plasma is shown in Fig. I.3.2-1.

Stabilization of RWM by the sector coils has been analyzed by the VALEN code in collaboration with Columbia University [3.2-1]. The equilibrium parameters are chosen as R₀/a=2.95 m/1.05 m, κ₉₅=1.79, δ₉₅ = 0.41, q₀/q₉₅/q_min = 4.28/7.52/2.1, l_i(3) = 0.63, p(0)/<p(ρ)> = 2.77 for the plasma of β_N = 4.5. The result is shown in Fig. I.3.2-2. In the figure, a curve marked with open squared corresponds to the passive stabilization only with the stabilizing plate. And the beta limit is β_N = 3.13. On the other hand, a curve marked with closed circles corresponds to the ideal wall limit. The beta limit is near 4.5. Other three lines correspond to cases with the sector coils feed-back with different three proportional gains. The result indicates that RWM can be stabilized up to β_N = 4.35.
Stabilization of NTM by the ECRF has also been investigated [3.2-1]. It was evaluated in a plasma based on the ITER-like plasma of I_p = 3.5 MA at B_t = 2.4 T with β_N = 1.8 but with reduced density (32% of nGW). The strength of the toroidal field was scanned. In the scan, Ip, ne and Te were changed in order to keep the safety factor, the Greenwald density fraction and β_N. For a simple evaluation, amount of the locally driven EC current was compared to the bootstrap current at the q = 1.5 location, which corresponds to the m/n = 3/2 NTM resonance. It was found that with expected injection power of 3 MW for the 110 GHz ECRF and of 4 MW for the 140 GHz ECRF, the local driven current density exceeds the boot strap current density in a wide range of B_t from 1.6 to 2.4 T. This roughly indicates that the ECRF powers are enough to suppress the m/n = 3/2 NTM in this range of B_t. More detailed analysis using TOPICS at 2.1 T was also carried out. The result indicates that the m/n = 3/2 NTM can be stabilized with less than 2 MW of the 140 GHz ECRF and roughly agrees with the analysis mentioned above.

JT-60SA has both upper and lower divertors. The lower divertor is planed to be compatible the ITER-like equilibriums and similar to that foreseen in ITER [3.2-1]. That is vertical target plates and a V-shape corner at the outer divertor. Function of this ITER-like divertor has been investigated by the SOLDOR/NEUT-2D code. It is confirmed that with "V-shape corner", a dense plasma that is similar to one obtained without "V-shape corner" but with gas-puff can be formed at the outer divertor. And with "V-shape corner" and gas-puff, detachment can be induced and heat flux to the hit point can be reduced greatly. The results indicate effectiveness of "V-shape corner". It should be noted that another simulation indicates that by increasing X-pint height, attached divertor can be retrieved with the same gas-puff. These results indicate that with "V-shape corner", divertor detachment can be controlled.

References
3.2-1 Fujita, T., et al., Proc. of the 21^st IAEA Fusion Energy Conference (Chengdu, 2006) CD-ROM file FT/P7-4.

JT-60SA has been designed aiming at the maximum utilization of the existing JT-60 facilities and hardware, such as buildings, plasma heating and current drive systems, some power supplies, diagnostics and cooling systems [3.1-1].
The cross sectional view of the tokamak is shown in Fig. I.3.3-1. A plasma aspect ratio down to about 2.6 was chosen in order to be able to survey optimum plasma shapes in view of the most cost effective DEMO reactor design. The major components of the tokamak will be installed into the inner space of a cryostat that has a diameter of about 14 m. The present design of divertor geometry was optimized to produce both high triangularity double-null/single-null plasmas and ITER-shape single-null plasmas within the same machine. A semi-closed vertical divertor with dome, rather than the ITER divertor configuration, was adopted to keep higher flexibility of plasma shaping capability.
The heating and current drive (H&CD) systems in JT60-SA will allow both power densities in excess of those expected in ITER as well as the capability to independently control heating, current and rotational profiles, a feature deemed as essential in order to be able to expand the understanding of advanced tokamak regimes. The high power plasma heating systems also bring about a substantial increase of the DD neutron yield to about 2×10¹⁹/shot. In order to increase the nuclear shielding effectiveness, boric acid water will be introduced into the double walled space of the vacuum vessel. This will help to reduce the nuclear heating in the TF and EF coils under a narrow shielding space constraint. Boron doped concrete will also be filled between the double wall of the cryostat for bio-shielding of the torus hall and prevention of the air activation. Owing to the expected high dose rate inside the vacuum vessel, remote handling system will be necessary, and will be developed for the maintenance and repair of in-vessel components such as divertor modules and first wall.
Figure I.3.3-2 shows the layout plan of the JT-60SA torus hall with all the major components. Four large ports have been devoted to remote handling. In addition, four beam towers of perpendicular injection positive ion based NBI (P-NBI), two beam tanks of tangential injection P- NBI, and a negative ion based NBI (N-NBI) located in the assembling hall will be reused. The waveguides of ECRF will be introduced from the upper neighboring room. Current lead boxes, valve boxes and other major equipments will be located nearby the cryostat. Plasma discharge operation is now planned to occur mainly during day time, as the regeneration of the divertor cryo-panels will be required every night in order to cope with plasma high density operation [3.3-1].

The superconducting magnets in JT-60SA consist of 18 toroidal field (TF) coils and 11 poloidal field (PF) coils, as shown in Fig. I.3.3-1. A NbTi superconducting conductor has been chosen for the TF coils owing to the limited maximum field strength of ~6.4 T. The choice of the final conductor design will be made on the basis of superconducting stability, manufacturing feasibility, optimal compatibility with the power supply, as well as cost.
PF coils consist of the Central Solenoid (CS) made of four identical coils (CS1, CS2, CS3, CS4) and seven equilibrium field (EF) coils (EF1-7). NbTi superconducting conductor has been chosen for all EF coils, including the divertor coils (EF3 and EF4), as their maximum field strength at the conductor surface of the EF coils is limited to ~6.2 T. Two kinds of NbTi conductor designs are expected for the EF coils, because the maximum magnetic field strength is quite different between the outer ring EF coils (EF1, EF2, EF5, EF6, EF7) with a large diameter, and the inner ring divertor coils (EF3, EF4).
For the CS a Nb₃Sn conductor has been chosen as a consequence of the maximum field strength (~9 T) needed to ensure a flux swing capability of about 40 Wb - necessary to sustain the flattop period of ~100 s with the rated plasma current of 5.5 MA under hybrid operation.
All TF coils are planned to be cooled down and cold-tested in advance to shipment to verify their functionality. Similar tests are also planned for the CS modules. The Japanese High Pressure Regulation Law will be applied for these superconducting magnets.

The vacuum vessel (VV) is a torus-shaped and double wall structure. The vacuum vessel is made of double-walled 316L stainless steel with low Co content (less than 0.05 wt%) to minimize induced activation. Filling up of the double-wall vacuum vessel with boric acid water is planned to enhance the neutron shielding capability of the vacuum vessel. A gravity support will be attached every 40 degree section with a pack of spring plates at the bottom of the vacuum vessel. Connecting plates between the neighboring vertical ports and/or gravity supports are introduced to increase the stiffness of the vacuum vessel against the twisting moment and seismic forces.
An 80 K thermal shield will be placed between the vacuum vessel and the cold structures of the superconducting magnets. The baking of the vacuum vessel, at 200C, will be done by means of heated nitrogen gas. Plasma operation during VV baking is not being planned because the neutron shielding capability would be lost during the baking using nitrogen gas. In addition, it would not acceptable from the view point of the heat load to the cryogenic system. Also it must be taken into account that, during plasma operation, the temperature of in-vessel components such as divertor cassette will be limited to around 60C.

The major in-vessel components of JT-60SA are the divertor cassettes, the inboard first wall, the stabilizing baffle plate, the fast plasma position control coils, and the sector coils for RWMs control.
The upper and lower divertor cassettes are asymmetric to allow ITER-shape plasmas and high triangularity plasmas. In the presently foreseen design, a mono-block type CFC divertor armor will be used for the outboard side to withstand a heat load of ~15 MW/m², while a flat tile type divertor armor is usable for the inboard side owing to the smaller expected heat load.
The CFC divertor armor will be connected to the divertor cassette in a way compatible with remote handling requirements. The cryo-pump can be installed behind the divertor cassettes in order to increase the pumping capability. On the basis of divertor performance simulations, the location of the cryo-panels will be finally decided considering the core and SOL plasma performances needed to contribute to the ITER experiments and to DEMO reactor design.
Two in-vessel field coils will be used for fast plasma position control. In order to control vertical and horizontal field at the same time using two coil blocks, independent power supplies will be connected to each coil block. The ampere turn of the fast plasma position control coils is ~80 kAturn. To suppress the excessive current and electromagnetic forces during plasma disruptions and VDEs, the insertion of external inductors in series to the coils and power supplies may be required.
The primary function of the stabilizing baffle plate is to enhance the ideal beta limit and to improve plasma positional stability. The stabilizing baffle plates are baked up to 200C as well as the VV. They have been designed as a double-wall structure made of SS316L and covered with bolted first wall.
The 18 sector coils for the RWM control will be installed around the openings of the stabilizing baffle plate to minimize the control response time. Since each sector coil has its own current leads, the magnetic field mode of m/n = 3/1 or 3/2 can be controlled depending on the configuration of the power supply connections.

The vessel body of the cryostat is a spherical shaped structure of about 14 m diameter. The cryostat vessel has a big top plate shifted up by about 1 m to maximize the inner space for the assembly and maintenance of cryogenic devices such as in-cryostat feeder, insulation break and TF terminal joints. The cryostat vessel is made of double-walled SS304 (Co < 0.05wt%) filled with boron doped concrete. The inner skin has a structure to support the weight of all the ports and to withstand the vacuum pressure. Since the role of the outer skin is just to contain the boron doped concrete, it is much thinner in comparison with the inner skin. Also, thin outer metal skin has the advantage to decrease the activation level outside the cryostat. Boron will efficiently reduce thermal neutron flux and hence the activation of Ar in the torus hall atmosphere. The gravity support made of SS304 (Co < 0.05wt%) is designed as a robust structure to support the dead weight of superconducting coils and the VV, including in-vessel components, operational electromagnetic loads, thermal stress during baking and standard operational conditions, and seismic loads.

The large foreseen annual neutron fluence of ~4x10²¹ neutrons/year will prevent human access inside the vacuum vessel after extensive experimental campaigns. Therefore, most of the in-vessel components will have to be compatible with remote handling requirements. The remote handling system is planned to use a vehicle type system, such as the one adopted in ITER for its shielding blanket. Four large horizontal ports are used to extend and support a rail and bring in/out the maintained components as shown in Fig. I.3.3-2. The functions of the remote handling system of JT-60SA are as follows: repair and exchange of divertor cassettes, repair of first wall for the inboard side and the baffle plates, while in-vessel coils and stabilizing baffle plates are assumed as permanent components. The weight of one divertor cassette is expected to be about 500 kg, while one first wall element is only a few kg. The remote handling system will have two types of headers to handle these two different loads.

References
3.3-1 Matsukawa, M., et al., Proc. of the 21^st IAEA Fusion Energy Conference (Chengdu, 2006) CD-ROM file FT/P7-5.

JT-60U was assigned as a core national device for joint research by the Nuclear Fusion Working Group of the Special Committee on Basic Issues of the Subdivision on Science in the Science Council of MEXT in January 2003. Using the JT-60 tokamak and other facilities, JAEA has performed research collaboration with the universities and NIFS. Accordingly, the joint experiments on JT-60 between JAEA and the universities by assigning university professionals as leaders of research task forces has been successful. The numbers of collaborators and research subjects had increased significantly since FY 2003 as shown in Fig. I.4.1-1.
In FY2006, 165 persons participated, who came from 25 research organs in Japan. Each subtheme is occupied by two leaders, one from university or NIFS and one from JAEA. The number of research subjects of the joint research was in total 33 in FY 2006, categories of which are shown in Table I.4.1-1. Fourteen journal papers and 23 papers in conference proceedings were published as a result of the joint research in FY 2006.

Status of collaborative research based on the IEA Implementing Agreement on cooperation on the Large Tokamak Facilities is described, following the annual progress report 2006 of the executive committee, which was presented at the 36th Meeting of the Fusion Power Coordinating Committee on Feb 27-28, 2007. This report covers from June 2005 to June 2006. In this period, there were personnel exchanges among the three Parties. The number of personnel exchanges, to which JAEA relates, is 8 in total. 3 personnel exchanges are from JAEA to EU, 1 from EU to JAEA, 0 from JAEA to US, and 4 from US to JAEA.
The exchanges from EU relate to the pedestal study titled "Participation to JET Ripple Experiments". After installation of ferritic steel tiles (FSTs) in JT-60U to reduce the toroidal field ripple, an improvement of pedestal performance was observed using large plasma configuration, where the ripple amplitude without FSTs was ~2%. In addition to this, previous JET/JT-60U similarity experiments indicated that the level of ripple amplitude is one of the differences between two devices. Therefore, to investigate in detail the role of toroidal field ripple on pedestal energy and ELM characteristics as well as the study of rotation effects, the "TF ripple campaign" including dedicated experiments using matched shape to the JT-60U plasma cross section was performed in JET with participations of N. Oyama and H. Urano. When the ripple amplitude increased, the pedestal density tends to decrease as well as the reduction of toroidal plasma rotation in co-direction to the plasma current. The change in the ELM characteristics was more obvious, where ELM frequency become higher as ripple amplitude increased and/or co-toroidal rotation decreased. This result is qualitatively similar to JT-60U. Detailed analysis is underway for the explanation of difference of pedestal structure between two devices.
The exchanges from US relate to the NNBI study titled "Experiments on negative ion systems". Dr. Grisham made three trips to collaborate with the JT-60U negative ion beam group during the period covered by this agreement. During this period, this collaboration was primarily engaged in developing techniques to more reliably steer the many beamlets which comprise a beam, and in finding ways to compress the beam envelope. This will be incorporated into the upgraded accelerators for JT-60SA, and will be essential in designing the ITER beam accelerators. We also studied processes precipitating breakdown in the accelerator column, and the evolution of oxygen contamination in the source plasma.
From April 2006 to March 2007, 4 papers were published in journals and 3 papers in conference proceedings, as a result of the IEA Cooperation among Large Tokamak Facilities related to JT-60U.