I. JT-60 Program

1. Experimental Results and Analyses

The JT-60U tokamak project has focused on the physics and engineering issues for the establishment of burning plasma operation in ITER and steady-state high β operation toward JT-60SA and DEMO. Since the optimization of current profile and pressure profile is essential for the stable and steady plasma operation, enhancing the control capability of these parameters is quite important. In addition, plasma rotation has a critical influence on magneto-hydro-dynamic (MHD) stability in the high β plasmas, edge pedestal performance and edge localized mode (ELM) characteristics. Thus, to expand the flexibility of heating and rotation profile control, the power supply system for 3 neutral beams (NBs) has been upgraded so as to enable the maximum injection period of 30 s in 2007. By making full use of the state-of-the-art tools of feedback control, heating/current drive and diagnostics, significant progress has been made in the integrated research on steady-state operation, transport, MHD stability, and edge pedestal physics.

1.1 Extended Plasma Regimes

Operation regime was extended toward the long sustainment of high βN with good confinement [1.1-1]. Real-time control of current profile was applied to high β plasmas, and the effectiveness of controlling the minimum in the q profile (qmin) in high β plasmas was clearly demonstrated [1.1-2]. Driven current profiles for the off-axis tangential NBIs were directly evaluated using motional Stark effect (MSE) diagnostics, and compared with ACCOME calculation [1.1-2].

1.1.1 Long Sustainment of High βN Plasma

Fig.I.1.1-1 Toward the development of the ITER hybrid operation scenario where the large current (low q95) operation for a long period is required both with high βN and a moderate current drive (CD) fraction, JT-60U has extended the operation regime, making full use of the increased heating power for a long pulse in this campaign; see Fig. I.1.1-1. Power supply systems for 3 units of NBs (about 6 MW) are modified to enable 30 s operation (formerly 10 s). The sustained period with βN=2.6 has been almost tripled from 10 s to 28 s in the high-βp ELMy H-mode plasma at low q95=3.2 (Ip=0.9 MA, Bt=1.5 T). The duration corresponds to about 16 times of the current diffusion time (τR). Good confinement (HH98y2≥1) is kept for 25 s, where the period is limited by the degradation of confinement (after t=29 s in Fig. I.1.1-1(a)) due to increase in density caused by enhanced recycling. In this discharge, off-axis bootstrap current maintains the flat safety factor profile at qmin~1, and the values of bootstrap current fraction (fBS) and non-inductively driven current (fCD) are 0.43 and 0.48, respectively. Weak and infrequent sawtooth activities were observed, but they did not affect the confinement. The onset of the neoclassical tearing modes (NTMs) was successfully prevented by the optimization of pressure gradient at low-q rational surfaces (3/2 and 2) through adjustment of the heating profiles.

1.1.2 Real Time Current Profile Control

Fig.I.1.1-2 The minimum value of the safety factor profile (qmin) affects the MHD instability related to the low-q rational surface, such as m/n=3/2 and 2/1 NTMs, where m and n are the poloidal and toroidal mode numbers, respectively. In order to demonstrate the effectiveness of qmin control on elimination of MHD activities, the real-time control of qmin was applied to the high-βp mode plasmas. Figure I.1.1-2 shows the waveforms of the discharge at Ip=0.8 MA, Bt=2.4 T (q95=5.4). Neutral beams (NBs) of about 14 MW were injected to produce high β plasma. When the diamagnetic stored energy Wdia reached 1.55 MJ (βN=1.7, βp=1.5), the m/n=2/1 NTM appeared, leading to decrease in Wdia by 22 %. In this discharge, we expected appearance of the m/n=3/2 NTM, so we set reference qmin,ref=1.7 intending to eliminate q=1.5 rational surface in the plasma. The lower hybrid (LH) waves were injected after t=7.5 s. The qmin control by LH power (PLH) starts at t=8 s, with primary parallel-refractive-index N//~1.65. When the qmin control started, PLH increased to raise qmin. The qmin reached to 1.7 at t=10 s, and the LH power decreased by the control. The qmin overshot the reference qmin,ref, and reached to 2 at t~11 s. At this timing, the magnetic and density fluctuations at 1.6 kHz were suppressed, and Wdia started increasing back to the initial value. Due to the reduction of LH power from t~10 s, qmin decreased down to qmin,ref=1.7 at t~12 s, showing that LH current drive was actually effective in sustaining the qmin value.

1.1.3 Off-Axis NBCD

Off-axis current drive is essentially important in advanced operation in ITER and JT-60SA. Although off-axis current drive by neutral beams (NBCD) is a candidate for the off-axis current driver, the characteristics of the driven current profile had not been investigated yet due to the difficulty in measuring its broadly distributed current density profile. Using the MSE diagnostics, the NBCD profile was measured in plasmas with Ip=0.8 MA and 1.2 MA at Bt=3.8 T [1.1-2]. In both cases, no MHD activity was observed except ELMs, and the spatially localized NB driven current profile was measured for the first time. The total amount of the measured driven current agreed with calculations by the ACCOME code in both cases. In addition, the measured driven current profile was consistent with neutron-emission profile measurement representing beam ion profile. However, the measured driven current density profile was more off-axis than that in the calculations.

1.1-1 Ide, S., the JT-60 team, Proc. 35th EPS Conf. on Plasma Physics, ECA 32F (2008) I1.007.
1.1-2 Suzuki, T., et al., Nucl. Fusion 48, 045002 (2008).

1.2 Heat, Particle and Rotation Transport
1.2.1 Toroidal Momentum Transport and Rotation Profile in L-mode plasmas

Recent tokamak studies have been emphasizing that the plasma rotation profile plays essential roles in determining confinement and stability. In order to establish a method for controlling the rotation profile, construction of physics basis of momentum transport and its effect on rotation profiles is required. Concerning toroidal momentum transport, most of the previous works evaluated the diffusive term utilizing the steady state momentum balance equation. However, it is recognized that the measured toroidal rotation velocity (Vt) profiles cannot be explained by the momentum transport coefficient evaluated by the steady state equation.

Parameter dependences of the toroidal momentum diffusivity (χφ) and the convection velocity (Vconv), and the relation between χφ and heat diffusivity (χi) are systematically investigated in typical JT-60 L-mode plasmas using the transient analysis by using the momentum source modulation [1.2-1]. Experiments have been carried out to investigate the momentum transport. The absorbed power is varied from 2.4 to 10.7 MW under otherwise similar conditions (Ip=1.5 MA, Bt=3.8 T, q95=4.2, δ=0.3, Vp=74 m3). These plasmas stay in a low collisionality and small Larmor radius regime with ρpol*~0.03-0.06 and ν*~0.07-0.26. As shown in Fig. I.1.2-1, the momentum diffusivity increases with increasing the heating power, and the shape of χφ profile is nearly identical. The Vconv profile takes non-zero value and has a minimum value at r/a~0.6. The momentum diffusivity at r/a=0.6 roughly scaled linearly with the heating power in this data set. We have also investigated Ip dependence of χφ and Vconv. During this Ip scan (Ip=0.87, 1.5, 1.77 MA), one CO tangential NB and one PERP-NB are injected with a similar absorbed power Pabs=3.3-4 MW and other plasma parameters were Bt=3.8-4.1 T, δ=0.3 and =1.3-1.9x1019 m-3. The inverse χφ at r/a=0.6 increases linearly with Ip, and such improvement of the momentum confinement is confirmed by steady toroidal momentum profiles. It is also found that toroidal rotation velocity profiles in steady state can be almost reproduced by χφ and Vconv estimated from the transient momentum transport analysis at low β (βN<0.4) as shown in Fig. 1.2-2.

Fig.I.1.2-1      Fig.I.1.2-2

1.2.2 Role of Pressure Gradient on Intrinsic Toroidal Rotation

The toroidal rotation velocity profile is determined by the momentum transport, external momentum source and intrinsic plasma rotation. It is important to separately evaluate contributions of the intrinsic rotation and the external induced rotation, and to understand the mechanism responsible for the generation of intrinsic rotation. Although the progress in understanding the physics of momentum transport and rotation has been made experimentally and theoretically in worldwide, the characteristics of the rotation profile including the spontaneous term is not yet sufficiently understood. This is due mainly to the experimental difficulty in separating the non-diffusive term and the spontaneous term.

We have identified the intrinsic rotation, which is not determined by the momentum transport coefficients and the external momentum input [1.2-2]. The momentum transport coefficients, such as χφ and Vconv, can be obtained separately from the transient momentum transport analysis, and the Vt profile is calculated using these coefficients, the external torque and the boundary condition of Vt [1.2-1]. From this approach, we have identified roles of externally induced rotation and the intrinsic rotation on the measured Vt profiles. The heating power scan is performed both in L-mode (Ip=1.5 MA, Bt=3.8 T, q95=4.2, δ~0.3, κ~1.3-1.4) and in H-mode plasmas (Ip=1.2 MA, Bt=2.8 T, δ~0.33, κ~1.4) in order to investigate the effect of plasma pressure on the spontaneous rotation. The absorbed power is varied over the range 2.4 MW<Pabs<11 MW for the L-mode plasma discharges, and 4.8 MW<Pabs<10 MW for the H-mode discharges. Although the measured Vt profile agrees with the calculation in the region 0.45<r/a<0.65, the measured Vt deviates from the calculated one in the CTR-direction in the core region 0.2<r/a<0.45 as shown in Fig. I.1.2-3. The difference in Vt is observed in the region where such large pressure gradients are measured. A good correlation between the difference in Vt (i.e. Vt(calculation) - Vt(measurement)) and the pressure gradient is found during the heating power scan: ΔVt increases with increasing pressure gradient in all cases including L-mode, H-mode, CO-, and CTR-rotating plasmas as shown in Fig. I.1.2-4. These results indicate that the local pressure gradient plays the role of the local value of spontaneous rotation velocity.

Fig.I.1.2-3      Fig.I.1.2-4

1.2.3 Dependence of Heat Transport on Toroidal Rotation in Conventional H-Modes in JT-60U

Temperature gradients are a key element in driving turbulent convection and causing anomalous heat transport in plasmas. The property of the turbulence driven by temperature gradient is believed to be provided by a strong increase of heat conduction which sustains a self-similar profile when the temperature profile exceeds a threshold in the temperature gradient (TG) scale length. The significant role of edge pedestal structure in H-modes, which is affected by the ELM activities, has been observed in many devices as a boundary condition for the heat transport in the plasma core. In JT-60U, the core temperatures vary in approximately proportion to the temperatures at the shoulder of the H-mode pedestal in a wide range of accessible densities in the type I ELMy H-mode regime, suggesting the existence of profile resilience. In H-mode plasmas where the ion channel is heated sufficiently by the positive ion-based neutral beams (NBs), the heat transport at the plasma core has been considered to be imposed strongly by the existence of a critical scale length of the ion temperature gradient LTi.

Understanding the effects of toroidal rotation velocity on the physical processes determining the heat transport and the pedestal structure in H-modes is one of the key issues in recent tokamak research. In JT-60U and DIII-D, it has been observed that the energy confinement is improved with the toroidal torque (and the resulting rotation) in co-direction to the plasma current Ip by the tangential NB. The confinement improvement with co-NBI accompanies the enhanced plasma pressure at the top of the H-mode pedestal together with the reduction of ELM frequency in case of JT-60U [1.2-3]. However, an underlying physics mechanism of this confinement improvement is not yet clear.

In this study, relation between heat transport in the plasma core and toroidal rotation as well as characteristics of the pedestal structure were examined in conventional ELMy H-mode plasmas in JT-60U. Conducting the experiments on power scan with a variety of toroidal momentum source generating the plasma rotation direction to co, balanced and counter with respect to the plasma current, dependence of the heat transport properties in the plasma core on toroidal rotation profiles was investigated. Energy confinement improvement was observed with increase in the toroidal rotation in co-direction. Heat transport in the plasma core varies, while self-similar temperature profile in the variations of toroidal rotation profiles are sustained. Pressure at the H-mode pedestal is increasing slightly with toroidal rotation in co-direction. Thus, energy confinement enhanced with co-toroidal rotation is determined by increased pedestal and reduced transport, brought on by profile resilience. In other words, heat transport in the plasma core is mainly determined by the saturation of temperature profile and is not strongly influenced locally by toroidal rotation. As shown in Fig. I.1.2-5, when the pedestal temperature was fixed between the cases of co and counter-NBI by adjusting the plasma density, the identical temperature profiles were obtained in spite of totally different toroidal rotation profiles. In H-mode plasmas where the ion channel is heated dominantly by the positive ion-based neutral beams, the saturation of ion temperature gradient governs the heat transport in the plasma core. As a result, large increase in heat conduction imposes the resilient profile of ion temperature, under which local effect of toroidal rotation profile on the scale length of ion temperature gradient is very weak [1.2-4].


1.2.4 Comparisons of Density Profiles in JT-60U Tokamak and LHD Helical Plasmas with Low Collisionality

Fig.I.1.2-6 In order to understand particle transport systematically in toroidal plasmas, electron density profiles were compared in JT-60U tokamak and LHD helical plasmas with low collisionality. The neoclassical particle transport in a low collisionality regime is significantly different for tokamak and helical plasmas. In helical plasmas, the 1/ν regime exists, where the neoclassical transport is enhanced as being proportional to 1/ν (ν is collisionality) in non-axisymmetric helical plasmas due to the presence of helical ripples [1.2-5]. On the other hand, anomalous transport in both plasmas seems to be related with common physics in toroidal systems. Gyrokinetic analyses showed that the quasilinear particle flux driven by drift wave instabilities exhibits weak dependence on the magnetic configurations [1.2-6].

Figure I.1.2-6 shows dependence of density peaking factors on collisionality for JT-60U ELMy H-mode plasmas (Ip=1 MA and Bt=2 T) and LHD plasmas (Bt=2.8 T) with a magnetic axis (Rax) of 3.5 m and 3.6 m [1.2-7]. Note that scaling studies indicated that both of these plasmas have the same weak gyro-Bohm like confinement feature [1.2-8]. Here, the density peaking factor was defined by a ratio of the central electron density at r/a=0.2 to the volume averaged density. In this figure, the abscissa indicates an electron-ion collision frequency normalized by a trapped electron bounce frequency (ν*b). The normalized collisionality of unity indicates a boundary between the collisionless region and the plateau region in both tokamak and single helicity (where only a single helical Fourier magnetic component exists) configurations. Thus, ν*b is a good index for showing the collisionality range for comparison. Here, the value of ν*b was calculated using plasma parameters at r/a=0.5. The density peaking factor increases with decreasing ν*b in JT-60U. The dependence in the collisionless region of LHD for Rax=3.5 m tends to approach that in the collisionless region of JT-60U, although the dependence in the collisionless region of LHD for Rax=3.6 m is reversed compared with that in the collisionless region of JT-60U.

The collisionality dependence of the density profile in JT-60U could be related to the anomalous inward pinch. The collisionality dependence of the density profile in LHD for Rax=3.5 m could involve the anomalous inward pinch, because it has been shown that neoclassical transport in LHD in the plateau region becomes the less pronounced as the magnetic axis is moved the more inward [1.2-5]. Therefore, the collisionality dependence of density profiles in LHD for Rax=3.5 m might become similar to that in tokamak plasmas, because gyrokinetic analyses showed that the quasilinear particle flux driven by drift wave instabilities exhibits weak dependence on the magnetic configurations [1.2-6]. On the other hand, the collisionality dependence of density profiles in LHD for Rax=3.6 m could be related to an increase in the neoclassical outward flux, because the convective flux for electrons is dominated by the temperature gradient driven flux, which increases in outward direction with decreasing collisionality in the 1/ν regime, in the datasets used here.

1.2-1 Yoshida, M., et al., Nucl. Fusion 47, 856 (2007).
1.2-2 Yoshida, M., et al., Physical Review Letters 100,105002(2008).
1.2-3 Urano, H., et al., Nucl. Fusion 47, 706 (2007).
1.2-4 Urano, H., et al., Nucl. Fusion 48, 085007 (2008).
1.2-5 Murakami S., et al., Nucl. Fusion 42, L19 (2002).
1.2-6 Yamagishi O., et al., Phys. Plasmas 14, 012505 (2007).
1.2-7 Takenaga H., et al., Nucl. Fusion 48, 075004 (2008).
1.2-8 Yamada H. et al., Nucl. Fusion 45, 1684 (2005).

1.3 MHD Instabilities and Control
1.3.1 MHD Stability Analysis of the High β Plasma for the Resistive Wall Mode Experiments

Fig.I.1.3-1 A very low rotation threshold was identified in JT-60U experiments in an investigation of the critical rotation for stabilizing resistive wall mode (RWM) by changing the toroidal plasma rotation using different combination of tangential neutral beams (NBs) [1.3-1]. Note that magnetic braking using external coils is not applied in JT-60U. The identified critical rotation is Vt ~20 km/s and corresponds to 0.3 % of the Alfvén velocity at the q=2 surface, which is much smaller than the previous prediction in DIII-D with magnetic braking. It is important for the RWM experiment to check the stability limit of external kink-ballooning mode without wall (no-wall limit) and with wall (ideal-wall limit) because RWM is destabilized between the ideal-wall limit and the no-wall limit. The MHD stability analysis code MARG-2D has been developed and has a useful interface for the analysis of experimental results with realistic plasma parameters, e.g. plasma pressure and current profile and equilibrium [1.3-2].

By using the MARG-2D code, the no-wall limit and ideal-wall limit in the RWM experiments was evaluated, and the value of the normalized beta (βN) at the no-wall limit was βN=2.3. Since the experimentally achieved βN was about 2.8, it is sustained above the no-wall limit. Figure I.1.3-1 shows the plasma displacement of n=1 RWM for the plasma at the no-wall limit. Here, n is the toroidal mode number, and the displacement profiles with the poloidal mode number from 1 to 6 are shown in this figure. This mode has a global structure that contains internal and external components. Namely the peaks of m=2~4 components correspond to rational surfaces q=2~4. The components of m=1 and m>4 are non-resonant and externally resonant ones. Moreover, the amplitude of this mode at the low-field side is larger than that at the high-field side. This is consistent with the experimental measurements. By this numerical analysis, we can estimate the no-wall and ideal-wall limit using experimentally obtained profiles. By comparing with the numerical and experimental results, it is found that the experimentally obtained critical rotation is still low as βN increases toward the ideal wall limit. These results indicate that for large plasmas such as in future fusion reactors with low rotation, the requirement of the additional feedback control system for stabilizing RWM is much reduced.

1.3.2 Simulation of Neoclassical Tearing Mode Stabilization

In JT 60U experiments, active stabilization of neoclassical tearing mode with m/n=3/2 or 2/1 using electron cyclotron current drive (ECCD) has been extensively performed. In addition, simulation of the stabilization of NTMs was performed using the TOPICS code combined with the modified Rutherford equation. In the simulation, the coefficients in the modified Rutherford equation were determined by comparing with experimental data. The TOPICS simulations were found to well reproduce the behavior of NTMs in JT 60U experiments. By using these coefficients, prediction analysis on NTM stabilization was performed. In particular, effects of ECCD location and ECCD deposition width on stabilization were investigated in detail [1.3-3,4]. Figure I.1.3-2 shows magnetic island width of m/n=2/1 NTM for various ECCD location (ρEC) and EC-driven current (IEC). The value of the vertical axis is normalized by the island width before ECCD. For the reference case, IEC=12 kA, the values of ECCD deposition width, full width of saturated island and mode location are 0.08, 0.11 and 0.59 in the normalized minor radius, respectively. The ratio of EC-driven current density to bootstrap current density at the mode rational surface, jEC/jBS, is about 1. This condition is similar to that in JT 60U experiments. In this case, NTM can be completely stabilized when misalignment is within about half of the full island width. When the misalignment is comparable to the island width, NTM is destabilized, and the width becomes wider by about 20 %. Such destabilization is actually observed in JT 60U experiments. When EC wave power is doubled (~23 kA), the V-shape profile becomes wider, that is, larger misalignment becomes acceptable for complete stabilization. However, at the same time, the destabilization effect also increases. When EC wave power is further increased to 29 kA, while allowed misalignment does not increase, the destabilization effect further increases. This suggests that precise adjustment is required to avoid NTM destabilization and subsequent mode locking even when abundant EC wave power is available. In addition to ECCD location, ECCD deposition width is another factor affecting NTM stabilization. In general, narrower ECCD deposition width has stronger stabilization effect. Figure I.1.3-3 shows a region of complete stabilization in the space of IEC and δEC*. Here, δEC* is full width at half maximum of ECCD deposition profile normalized by saturated island width. As shown in this figure, the boundary for complete stabilization increases roughly with IEC0.5. Thus, stabilization effect is roughly proportional to IECEC2, showing that ECCD width is an important factor for NTM stabilization. The dependence is similar to the previously obtained result on m/n=3/2 NTM stabilization.

Fig.I.1.3-2      Fig.I.1.3-3

1.3.3 Instability in Ion Cyclotron Frequency Range [1.3-5]

Fig.I.1.3-4 The fluctuations in ion cyclotron range of frequency (ICRF) are driven by the presence of non-thermal ion distribution in magnetically confined plasmas. ICRF antennas are used as pickup loops for detecting electrostatic and/or electromagnetic fluctuations. Two sets of ICRF antennas, which are installed with the distance of 1.67 m in the toroidal direction, are used in this experiment and the toroidal wave number can be evaluated due to the small pitch angle of the toroidal magnetic field line. Two types of magnetic fluctuations are detected: one is due to high energy D-ions from neutral beam (NB) injections and the other is due to fusion products (FPs) of 3He and T-ions. These fluctuations have been reported as ion cyclotron emissions (ICEs) in the burning plasma experiments on large tokamaks [1.3-6]. In this experiment, the first measurement of the toroidal wave numbers of those spontaneously excited waves is described. The modes due to D-ions have zero or small toroidal wave number kz. On the other hand, the measurement of finite kz in the modes due to FP-ions supports the excitation of the Alfvén waves is the possible origin of FP-ICEs. Figure I.1.-3-4 shows a typical intensity plot of the fluctuation amplitude as a function of time and frequency. The time sequence of NBs is indicated in the figure. Two sharp peaks, of which frequencies are corresponding to the fundamental and 2nd harmonic cyclotron frequencies of 3He near the outer midplane edge of the plasma, appear when the tangential positive-ion-based NBs (P-NBs) are injected. After perpendicular P-NBs are injected, relatively broad peaks due to D-ions are detected. A peak with the lowest frequency below 10 MHz appears after negative-ion based NB is injected. This is considered to be due to T-ions. It is also observed that the excited modes due to FP-ions (3He and T-ions) have different characteristics: driven by different neutral beams and having different parameter dependence. ICE due to T-ions has no harmonics and the value of ω/Ωci is smaller than that due to 3He-ions. Both beam-driven ICEs and FP-ICEs are clearly observed, and those spatial structures are also obtained on JT-60U.

1.3.4 Development of Active MHD Diagonostic System [1.3-7]

Fig.I.1.3-5 In order to actively diagonose the MHD stability such as RWM, edge localized mode (ELM) and Alfvén eigenmode (AE), we have been developing an active MHD diagnostic system in JT-60U. This system has two one-turn coils located 180 apart toroidally in the JT-60U vacuum vessel. These coils are connected to bipolar power supplies that can apply ±60 V and ±60 A as the maximum voltage and current, respectively. According to the calculation by Biot-Savart method in vacuum, the maximum magnitude of radial magnetic perturbations is predicted to be δBr ~ 0.1 G at the plasma surface. Figure I.1.3-5 shows the Fourier components spectrum of magnetic perturbations produced by this system. In the case where the coil currents are out of phase and in phase with each other, the dominant mode number is odd and even, respectively. These toroidal mode spectra are fairly broadened up to the intermediate n~20. This enables us to perform the active MHD diagnostic for low-n (RWM and low-n AE) to intermediate-n (ELM and AE) MHD instabilities. We expect that these could lead to deeper understanding of MHD stability.

1.3-1 Takechi, M., et al., Phys. Rev. Lett. 98, 055002 (2007).
1.3-2 Aiba, N., et al., Comput. Phys. Commun. 175, 269 (2006).
1.3-3 Isayama, A., et al., Nucl. Fusion 47, 773 (2007).
1.3-4 Ozeki, T., et al., Phys. Plasmas 14, 056114 (2007).
1.3-5 Ichimura, M., et al., Nucl. Fusion 48, 035012 (2008).
1.3-6 Dendy R.O., et al., Nuclear Fusion 35, 1733-42 (1995).
1.3-7 Matsunaga, G., et al., Proc. 10th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems 2007 (Kloster Seeon, 2007) CD-ROM file P-10.

1.4 H-Mode and Pedestal Research
1.4.1 Effect of Toroidal Field Ripple and Toroidal Rotation on H-Mode Performance and ELM Characteristics in JET/JT-60U Similarity Experiments

Fig.I.1.4-1 In order to understand the effect of toroidal field (TF) ripple and toroidal rotation (Vt) on H-mode performance and ELM characteristics, dedicated ripple experiments were performed in JET and JT-60U using a matched plasma shape [1.4-1]. After the installation of ferritic steel tiles in JT-60U, toroidal field ripple, δr, near the outer midplane was reduced to ~0.5 % at Bt =2.2 T. In JET, on the other hand, δr can be varied by selecting the appropriate differential current between odd and even set of coils out of 32 TF coils, providing in this case four levels of ripple amplitude of δr=0.1, 0.5, 0.75 and 1 %. Although the same level of ripple amplitude was successfully obtained in both devices, the Vt in JET was still higher than that in JT-60U due to a different level of fast ion losses. Since a correlation between δr and VT was found in JET plasmas, it is difficult to separate the two parameters, δr and Vt, in JET experiments. A series of power and density scans indicated that plasmas with lower δr and/or larger co-Vt are favorable to achieve higher pped and HH98y2 factor in both devices. As for ELM characteristics, larger co-Vt seems to increase the ELM energy loss together with the reduction of the ELM frequency. Sustaining a high neped with type I ELMs was still difficult in JT-60U plasmas even with FSTs (δr ~ 0.5 %), while neped in JET plasmas at δr = 0.75 % could be increased to similar value as at δr = 0.1 % using the same amount of gas puffing, as shown in Fig. I.1.4-1. Therefore, there are still some differences between the two devices. In order to obtain a better prediction for ITER plasmas, further investigations into the effects of the TF ripple and the toroidal rotation are necessary, and the acceptable level of ripple amplitude in ITER is still uncertain from these experimental results in both devices.

1.4.2 Dimensionless Parameter Dependence of H-mode Pedestal Width Using Hydrogen and Deuterium Plasmas in JT-60U

In H-mode plasmas, the edge pedestal structure determines the boundary condition of the heat transport of the plasma core. Therefore, it is of primary importance to understand the physical processes determining the edge pedestal structure. The structure of the H-mode pedestal is composed of a height of the plasma pressure and a spatial width in which a steep pressure gradient is formed. In this region, the periodic expulsion of energy and particles is commonly observed due to the existence of the edge localized modes (ELMs) triggered by a steep pressure gradient and/or a large bootstrap current. However, the dependence of the spatial width of the H-mode edge transport barrier on local and global plasma parameters is not well understood. In particular, knowledge of the pedestal width Δped based on dimensionless parameters is of great help for the extrapolation towards a next step device. However a well-validated means of predicting the spatial width of the H-mode transport barrier in ITER is still lacking and remains a major topic of research. Attempts to project a pedestal width for ITER through physics based or empirical scaling from multi-tokamak databases are underway to extend turbulent models through the pedestal region in recent tokamak research.

A variety of empirical scalings of Δped during the well-developed type I ELMy H-mode phase have been proposed empirically, based on theoretical or semi-empirical models. However, these scalings vary from machine to machine and with the operational regime. In particular, a main argument on the pedestal width Δped is whether Δped scales as the normalized poloidal Larmor radius ρpol* or βpol. This disagreement of the scalings intermingling ρpol* and βpol can be caused by the existing edge stability boundary for ELMs. In this region, ρpol* and βpol are linked strongly and thus are hard to separate out in the standard dimensionless parameter scan using a single gas species. To distinguish these variables, a pair of experiments in hydrogen and deuterium plasmas was conducted in this study. Explicit difference between ρpol* and βpol is the mass dependence of ρpol* (∝ m0.5) in contrast with no mass dependence in βpol.

Both the database analysis and the dedicated experiments on the mass scan indicated that the pedestal width depends very weakly on the plasma particle species or ρpol*. Identical profiles of the edge Ti obtained in the experiments suggested that the pedestal width more strongly depended on βpol than ρpol* as shown in Fig. I.1.4-2. The experiment on βpol scan was also performed. Higher βpol plasma had higher pedestal Ti value accompanied by wider pedestal width in spite of almost identical ρpol* at the pedestal. Based on the experiments on the dimensionless parameter scan, the scaling of the pedestal width was evaluated as Δped ∝ ap ρpol*0.2 βpol0.5 [1.4-2].

1.4-1 Oyama, N., et al., "Effect of toroidal field ripple and toroidal rotation on H-mode performance and ELM characteristics in JET/JT-60U similarity experiments," to be published to Journal of Physics: Conference Series.
1.4-2 Urano, H., et al., Nucl. Fusion 48, 045008 (2008).


1.5 Divertor/SOL Plasmas and Plasma-Wall Interaction
1.5.1 Particle Control in Long-Pulse Discharges

Fig.I.1.5-1 The particle retention in the plasma surface wall has been investigated in a series of long-pulse discharges by a particle balance analysis [1.5-1]. As shown in Fig.I.1.5-1, the first 13 discharges with a line-averaged electron density lower than 3.5×1019 m-3, which corresponds to ~75 % of the Greenwald density, the particle retention in the first wall gradually decreased. This is due predominantly to outgassing from the divertor plates with an increase of the surface temperature. In contrast, in the high-density discharges with a line-averaged electron density of 75 - 83 % of the Greenwald density, the particle retention changed to increase since the wall-pumping rate becomes positive. Several reasons are candidates for the positive wall-pumping rate and investigated. First, with increasing electron density, the increase of the divertor plate temperature was suppressed, resulting in lower outgassing rate compared to that of the low-density discharges. Second, in the high-density discharges, the co-deposition of deuterium with carbon became significant to enhance the fuel retention rate. Third, in addition, wall-pumping process was enhanced dynamically with increasing particle flux.

1.5.2 Radiative Plasma Control

Reduction of heat loading appropriate for the plasma facing components is crucial for a fusion reactor. Power handling by large radiation power loss using impurity (Argon) gas seeding was studied to sustain the high confinement ELMy H-mode plasma with the large radiation power under the wall saturated condition, where particle recycling flux changes during the long discharge. Controllability of the large radiation in the good energy confinement plasma (HH98y2=0.78-0.87) was investigated by the radiation feedback control of the Ar gas puff rate, using the bolometer signals viewing the main plasma edge. Total radiation fraction of Prad/Pabs = 0.75-0.9 was maintained continuously during Ar gas puffing (up to 13 s so far) under the outgassing condition from the plasma facing components.

1.5.3 SOL Fluctuations and ELM

Characteristics of the ELM filaments and intermittent plasma events have been investigated from the fast sampling database of three reciprocating Mach probes measured in 2006. In particular, in-out asymmetry in the plasma propagation of these transient events was investigated [1.5-2, 3].

ELM plasma measurements showed two transient convective fluxes during the ELM event: one was the large, short convective flow caused by the filaments, and the other was the flow reversal over a wide SOL region. The latter fact was observed only at the High-Field-Side (HFS) SOL. During the early period of type I ELM event as shown in Fig. I.1.5-2, seven large peaks in the ion saturation current (js) appeared only at the upstream side of the HFS SOL (jsHFS-up), and the radial propagation was not seen. On the other hand, at the Low-Field-Side (LFS) midplane, multi-peaks in js were seen both at the upstream and downstream sides of the Mach probe, and radial propagation with v=0.4-3 km/s was evaluated. In Fig. I.1.5-2, delays of the first few peaks were smaller than the parallel convective transport from the LFS midplane (~190 μs). This fact suggested that ELM filaments extend to the HFS edge and are ejected into the SOL. In the characteristics of the filaments at the HFS SOL, the separation of the maxima (τp-pmid) was 25-55 μs, which was larger than those at the LFS midplane. The toroidal mode number was estimated to be relatively large (n=18-44) from the toroidal rotation velocity measurement. After the appearance of the multi-peaks, the flow reversal of the SOL plasma occurred over a wide SOL region.

Intermittent convective plasma transport, i.e. "plasma blobs" has been observed in the LFS SOL, which is thought to play a key role for cross-field transport of the SOL plasma. Conditional averaging method was employed to reveal the typical burst's profile, and Figure I.1.5-3 shows the typical time evolution of js at the LFS SOL and the positive spikes have the common property of a rapid increase and slow decay. This feature was similar to that of theoretical prediction for plasma blobs. The conditional average of floating potential (Vf) measured simultaneously showed that Vf changes from positive to negative with respect to <Vf> as a large burst of js passes by. This result gives the internal potential structure of plasma blobs. The radial velocity of the plasma blob was estimated by using delayed time between the peaks of js and Vf to be about 0.43 km/s.

Fig.I.1.5-2      Fig.I.1.5-3

1.5.4 Spectroscopic Study in Divertor

Fig.I.1.5-4 The spectral lines of C2+ and C3+ emitted around the X-point in a detached plasma with MARFE were measured with a VUV spectrometer and a two-dimensional visible spectrometer in order to understand the controllability of the dominant radiation from C3+ [1.5-4]. As shown in Fig. I.1.5-4, it has been found that C3+ was produced by the volume recombination of C4+ and the ionization of C2+ comparably. In contrast, the volume recombination of C3+ was not detected, and the ionization flux of C3+ was less than 1 % of the C3+ generation flux. Thus, the C3+ generation flux was higher by two orders of magnitude than the loss flux. This result suggests that loss mechanism of large number of C3+ ions from the radiation dominant region such as parallel or cross-field transport is significant. Because the ionization flux of C3+ was much smaller than the recombination flux of C4+ around the X-point, the predominant source of C4+, which recombined into C3+, was presumably the main plasma. Since the flux of C4+ is determined by the transport in the main plasma, it is difficult to control the C4+ flux.

Similarly, significant ionization of C2+ into C3+ and no recombination of C3+ into C2+ indicated that the source of C2+ exists in the divertor region, which can be increased for instance by seeding CD4. Because the source rate of C3+ from the main plasma (the recombination of C4+) and the divertor (the ionization of C2+) was found to be similar as described above, the radiation loss control by impurity seeding from the divertor will work partly.

1.5.5 Study of Dust Dynamics

Movement and velocity of the dusts were measured with a fast visible TV camera (2-6 kHz) from tangential port to determine the trajectory and velocity of the dusts. In the main SOL, many dusts with various directions were observed, particularly, in the first shot after hard disruptions and overnight (3-7 hours) He-GDC. Many dusts were ejected from the inner divertor, in particular, for the high strike-point case: large ELM heat and particle loading on thick deposition layers may enhance producing dusts due to large thermal stress. Velocity of the toroidal movement (0.2-0.5 km/s) was faster than that of the radial movement (<0.05 km/s). Toroidal movement was mostly towards the ion drift direction (Ip), which is consistent with the SOL flow measurement at HFS and LFS SOLs.

1.5.6 Progresses of SOL and Divertor Code Development and Simulation Study

In order to understand the divertor pumping for particle control, the roles of atomic and molecular processes of ionization (Ie), charge exchange (CX), elastic collision (EL), and dissociation (DS) were investigated by the SOLDOR/NEUT2D code, which coupled the 2D plasma fluid modeling code SOLDOR and the Monte Carlo neutral kinetic modeling code NEUT2D. Characteristics of the incident neutrals into the exhaust slot were investigated from the dependence of the neutral pumping efficiency on the strike point distance from the slot. A case of small distance (2 cm) showed high pumping efficiency, where the most of neutrals produced on the target go toward the exhaust slot directly, except for the disappeared neutrals by Ie and DS processes. Whereas, for the long distance case of 10 cm, those neutrals were scattered at random by CX and EL and a few of them go toward the slot, leading a low pumping efficiency. Atomic and molecular processes was also identified when the divertor plasma changed from low ne /high Te (ne-spave=0.7×1019 m-3 and Te-spave=60 eV) to high ne /low Te (ne-spave=21×1019 m-3 and Te-spave≤1 eV). The incident neutrals to the slot for the low ne /high Te case were dominated by atoms produced through DS, then CX and EL. Molecules due to EL were are dominant for the high ne /low Te case [1.5-5].

A kinetic neutral model and a fluid neutral model in the divertor code have been compared under the detachment condition through collaboration with KEIO University. The SOL radial profiles of basic plasma and neutral parameters with the fluid neutral model were fitted to those with the kinetic model as the first step of the comparison, where the drift effects were also introduced. In the SOL, no significant effect of the drifts on the radial profiles of the plasma temperature and density was seen. However, in the divertor region, large differences were seen in the two models without the effects of drift, and drift effects tended to enlarge the difference. As a result, improvement of the fluid neutral model is required in the divertor region. [1.5-6].

1.5.7 Tungsten Erosion and Deposition

Tungsten tiles have been installed at the upper row of the outer divertor in the No.8 port section since 2003. Poloidal and toroidal distributions of tungsten deposition on the divertor CFC/graphite tiles were investigated. A neutron activation method was used for the absolute measurement of W density by the 186W(n,γ)187W reaction by slow neutrons in JAEA/FNS (Fusion Neutronics Source). Conventional surface analysis methods such as EDX (Energy Dispersive X-ray spectrometry) and XPS (X-ray photoelectron spectroscopy) were also used.

For the divertor tiles exposed in 2003-2004, tungsten deposition on the dome tiles was found only near the top surface (within depth of a few μm), while tungsten on the inner divertor tiles was co-deposited with carbon to the depth up to about 60 μm. The neutron activation method could measure tungsten in these thick co-deposition layers. Ion beam analysis such as PIXE (Particle Induced X-ray Emission) was not appropriate for this case since only near-surface tungsten (~10 μm) was precisely measured by this method. It was found that tungsten deposition profile was not uniform toroidally. Tungsten surface densities near the inner strike point and on the outer wing were much higher at the toroidal position of 0 degree (near the W-tile position) than those toroidally separated at 60 degree (~3 m). Here the toroidal angle was defined as the angle viewed from the top of the torus (counter-clockwise).

Detailed measurement of the W toroidal distribution on the outer wing showed significant localization near the W-tile section. Since ionization length of W atom in the divertor plasma was more than an order of magnitude shorter than the distance of ~3 m, this long-distance deposition on the outer wing was not attributed to the local deposition of sputtered tungsten atoms. This result indicated that transport of tungsten ions in the outer divertor plasma was significantly affected by the cross-field transport in the private flux region. Toroidally localized W deposition was also observed on the inner divertor tile, thus significant amount of the tungsten could be also transported to the inner divertor through the private plasma by the cross-field transport such as drifts. At the same time, effect of the plasma flow along magnetic fields on the tungsten transport was also investigated.

1.5.8 Hydrogen-Isotope Retention

At the outer board first wall graphite tiles, it has been found that the deuterium retention was higher than the hydrogen retention even after hydrogen discharges for tritium degassing in 2004 [1.5-7]: at the surface of the tiles (< 0.1 μm), the hydrogen retention was higher than the deuterium retention. In contrast, the deuterium retention was higher in the deeper range between 0.1 μm and 1 μm (corresponding to the implantation depth of the fast ion), which contributed largely to the total amount of hydrogen-isotope retention, although it decreased with the depth. In a further depth range between 1 μm and 10 μm, the deuterium retention decreased gradually and finally reached a very low ratio of D to C (D/C~10-4) [1.5-8].

The hydrogen-isotope retention in the side surfaces of the inner and the outer board first wall tiles was measured, and D/C was found to be comparable to that of the outer divertor tiles [1.5-9]. Compared to the tile surface of the inner divertor tiles, where the hydrogen-isotope retention was due predominantly to the co-deposition with carbon, D/C in the side surfaces of the first wall tiles was smaller. Since the total area of the tile side surface of the first wall tiles is much larger than that of the inner divertor tiles, the contribution of the tile side is under investigation. The deuterium depth profile was similar to that of boron, indicating that the deuterium was incorporated with boron.

The deuterium depth profile of the divertor tiles without the degassing discharges at the end of 2005-6 operations was measured for the first time [1.5-10]. Comparison of the deuterium depth profile at the inner divertor tile surface with and without the degassing discharges showed that about 90 % of the trapped deuterium was removed by the hydrogen degassing discharges. However, in the range deeper than ~1 μm, the deuterium retentions with and without degassing discharge were comparable within a factor of 2 (except for the dome top tile). Furthermore, it has been found that the deuterium concentration in the deep area (up to 16.4 μm) was not decreased. A similar depth profile was also found in the divertor tiles of ASDEX-U, and it was concluded that this is due to permeation of hydrogen isotopes along the orientation of carbon fibers [1.5-11]. This is also a candidate mechanism to explain the wall-pumping in the series of the high-density and long-pulse discharges [1.5-1].

1.5-1 Nakano, T., et al., Nucl. Fusion 48, 085002 (2008).
1.5-2 Asakura, N., et al., "ELM propagation in the low- and high-field-side Scrape-off Layer of the JT-60U tokamak," to be published in J. Phys.: Conf. Ser. (2008).
1.5-3 Asakura, N., et al., "Application of Statistical Analysis to the SOL Plasma Fluctuation in JT-60U," submitted to J. Nucl. Mater.
1.5-4 Nakano, T., et al., "Radiation process of carbon ions in JT-60U detached divertor plasmas," submitted to J. Nucl. Mater.
1.5-5 Kawashima, H., et al., Contrib. Plasma Phys. 48, 158 (2008).
1.5-6 Hoshino, K., et al., Contrib. Plasma Phys. 48, 136 (2008).
1.5-7 Yoshida, M. et al., "Hydrogen isotope retention in the first wall tiles of JT-60U," submitted to J. Nucl. Mater..
1.5-8 Hayashi, T. et al., J Nucl. Mater. 363-365, 904 (2007).
1.5-9 Nobuta, Y. et al., "Retention and depth profile of hydrogen isotopes in gaps of the first wall in JT-60U," submitted to J. Nucl. Mater.
1.5-10 Hayashi, T. et al., "Deuterium depth profiling in graphite tiles not exposed to hydrogen discharges before air ventilation of JT-60U," submitted to J. Nucl. Mater.
1.5-11 Roth, J. et al., J Nucl. Mater. 363-365, 822 (2007).

2. Operational and Machine Improvements
2.1 Tokamak Machine
2.1.1 Development of Supersonic Molecular Beam Injection

In order to control plasma pressure at the pedestal region and to evaluate the effect of fuel on the self-organization structure of plasma, a supersonic molecular beam injection (SMBI) system has been installed in JT-60U device. The SMBI system has been collaborated with CEA/Cadarache in France. The schematic diagram of the SMBI is shown in Fig. I.2.1-1. The nozzle heads were put on the wall of vacuum vessel (VV).

However, in the first SMBI system, the seal of the nozzle head was evaporated during baking of the vacuum vessel at a temperature of 300 degree C, causing a large seal leak. To improve this, heat-resistant materials were tested at JAEA and CEA/Cadarache, and FLUORITZ-HR was confirmed as a most promising seal material that showed no evaporation at temperature lower than 280 degree C. Thus, FLUORITZ-HR was adopted as a new seal material, and the baking temperature was decreased from 300 degree C to 280 degree C and gas pressure of the nozzle was increased from 0 MPa to 0.5 MPa to reduce the possibility of leak due to deformation of the seal. After these countermeasures, the SMBI successfully worked without vacuum leak.

Time evolutions of the plasma parameters in the case of high field side injection of SMBI are shown in Fig. I.2.1-2. The frequency of the SMBI is 10 Hz, and the pressure of SMBI is 0.3 MPa. The fast response of the plasma density synchronizing with the movement of the nozzle head of SMBI is confirmed.

Fig.I.2.1-1      Fig.I.2.1-2

2.1.2 Break-away of the Carbon Tiles

I.2.1-3 A carbon tile in the outer dome of the divertor was broken away and the tile dropped into the vertical port in June, 2006. The cause of the break was considered to be due to the heat flow from a plasma to the carbon tiles. The trace of heat concentration was observed at the bolt and nut between the tile and the basement of tile. The nut was melted by the heat flow and the tile was broken away from the basement.

Countermeasures for this break-away were taken during the maintenance period from May to September in 2007. In the inspection just after the break-away, bolt holes with unprocessed edges, which made a slight gap between the tile and the basement to reduce heat conduction between them, were found commonly for some of the dome tiles. Therefore, all bolt holes of the dome basements were processed. In addition, carbon sheets were set between the dome tiles and basements to ensure heat conduction though the dome had originally been designed as components that were not expected to directly receive high heat flux. However, the break-away of the carbon tile in the outer dome was found in March, 2008 again. The position of the tile is the same as that of the broken-away tile in June, 2006. The circumstance of the broken-away tile is shown in Fig. I.2.1-3. All tiles in outer dome (125 tiles) were checked again. No loosened bolt and melting were found except the broken-away tile. It is considered that the countermeasures for the first break-away was effective except the broken-away tile, although the cause has not been understood perfectly. The same countermeasures were taken again for the broken-away, and the position of the outer separatrix leg has been controlled at more than 0.5 cm away from the dome since this event.

2.2 Control System
2.2.1 New Real-Time Control Functions for Advanced Plasma Operation

For efficient exploration toward high performance plasmas in JT-60 operation, more advanced real-time control functions are required to be applied to the experiment. The major developments conducted in 2007 are presented as follows: (a) New real-time control of ion temperature (Ti) profile and toroidal rotation velocity (Vt) measured by the fast charge exchange recombination spectroscopy (CXRS) system during neutral beam heating [2.2-1]. (b) The integral control method newly employed for the real-time control of radiation power profile in a divertor plasma. (c) A digital filter newly applied to the separatrix strike position on the divertor dome plates for stable strike point control. (d) Multiple integral controls applied to a single real-time profile control. (e) Prompt plasma shape evolution viewer newly developed for efficient evaluation before the creation of JT-60 database using the real-time plasma shape visualization system [2.2-2, 3]. (f) Voltage control newly developed for safe operation even in case of deteriorated electrical insulation of the PF coil power supply. These new advanced plasma real-time control functions have contributed to achievement of annual mission parameters specified in JT-60 experimental plan.

2.2.2 Development of an FPGA-based Timing Signal Generator

I.2.2-1 The timing system in the supervisory control system of JT-60 [2.2-4] is composed of specially ordered CAMAC modules for limited usage [2.2-5]. This system generates trigger signals and a clock signal necessary for measurements and controls in the JT-60 experiment. Recently, we have been facing difficulties in cost-effective maintenance due to CAMAC standardized hardware, and in increasing troubles due to the aged deterioration.

The following design guidelines for a new system configuration were employed: (a) A master clock cycle should be changed from 1 kHz to 40 MHz to improve accuracy. (b) FPGA (Field Programmable Gate Array) was chosen to minimize delay time for the logical calculation, and enhance the flexibility for changing logic flows. (c) The addition and change of the timing signal should be easily conducted in the new system. (d) The timing signal should be transmitted to the subsystem in less than 1 μs by using the optical cable. (e) The number of cables should be minimized, which requires multiplex transmission. (f) The interfaces between the supervisory timing system and subsystem one should remain unchanged.

I.2.2-2 Figure I.2.2-1 shows a new timing system hardware configuration. The timing system supervisor is composed of the VME-bus system with the host computer through Ethernet network. Figure I.2.2-2 shows the circuit configuration of the timing signal generator. Response time measurement of timing signal transmission has been carried out by using two timing signal generators connected with optical fiber cable as shown in Fig. I.2.2-3. Test result is shown in Fig. I.2.2-4 [2.2-6]. The elapsed transmission time was 276.5-326.5 ns, and reading time from input register was 50-100 ns, and writing time to output register was 62.5 ns. The optical transmission time takes 333 ns between supervisory timing system and subsystem. Total sum of test result and optical transmission time take 600-650 ns in signal transmission while 30-40 μs is needed in the original CAMAC system. This value is less than allowable delay time 1 μs.

Fig.I.2.2-3      Fig.I.2.2-4

2.2-1 Yoshida, M., et al., "Real-time measurement and feedback control of ion temperature profile and toroidal rotation in JT-60U," submitted to Fusion Eng. Des.
2.2-2 Sueoka, M., et al., Fusion Eng. Des. 83, 283 (2008).
2.2-3 Sueoka, M., et al., Fusion Eng. Des. 82, 1008 (2007).
2.2-4 Kurihara, K., et al., Fusion Eng. Des. 81, 1729 (2006).
2.2-5 Akasaka, H., et al., Fusion Eng. Des. 71, 29 (2004).
2.2-6 Kawamata, Y., et al., Fusion Eng. Des. 83, 198 (2008).

2.3 Power Supply System
2.3.1 Operational Experience

Annual inspections and regular maintenances for the power supply system have been conducted to maintain high performance operation as shown in Table I.2.3-1. In addition, the renewal and special maintenances have been also conducted to avoid troubles for the aged-deteriorated system as shown in Table I.2.3-2. These activities contributed to achieving safe operation of the systems. In addition, the renewal and special maintenances have been also conducted to avoid troubles for the aged-deteriorated system as shown in Table I.2.3-2. These activities contributed to achieving safe operation of the systems.

Table.I.2.3-1      Table.I.2.3-1

2.3.2 Investigation of MG Shaft Vibration in the Motor-Generator for Additional Heating Devices

The motor-generator with 300-ton flywheel for additional heating devices (H-MG) has double layers of stator windings connected together to a single output through the AC circuit breakers.

A problem of the excessive shaft vibration was observed in the test operation with two-single-layer connection. Voltages between three phases for two single layers of stator windings are shown in Fig. I.2.3-1 (a) and (b). The voltages from the phase V commonly oscillated, while the voltage of U-W phases showed no oscillation.

When two single layers were connected together, no excessive shaft vibration occurred at all as shown in Fig. I.2.3-2. It could be understood that the V-phase stator winding might be somehow differently reconstructed from the original one in 2004.

Fig.I.2.3-1      Fig.I.2.3-2

2.4 Neutral Beam Injection System

In the campaign of 2007, the injection time of the negative-ion-based NBI (N-NBI) unit was successfully extended from previous 20 s to 27 s at 1.0 MW by modifying the negative ion sources. Four perpendicular positive-ion-based NBI units were also upgraded to extend the injection pulse length up to 30 s at 2 MW of injection power. These long pulse injections from the N-NBI and P-NBI units significantly contributed to the study on quasi-steady state plasmas in JT-60U.

2.4.1 Long Pulse Operation of NBI System

On JT-60U, there are 11 positive-ion-based NBI (P-NBI) units, each of which injects 2 MW D0 beams. Out of P-NBI units, 4 tangential P-NBI units have been already upgraded to inject the D0 beams for 30 s in 2006. In 2007, 4 perpendicular P-NBI units were additionally upgraded to extend the injection pulse length up to 30 s by mainly increasing the capacity of the power supplies. The injection pulse length of the upgraded P-NBI units is extended while the beam-limiter temperature increased by heat load of the high-energy re-ionized particles from the D0 beams was confirmed to be reduced to an allowable level. The pulse lengths of the upgraded P-NBI units are extended from the previous 10 s to ~ 20 s. By using the tangential units modified in 2006 and the perpendicular units, high-power and long-pulse D0 beams 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 injection pulse length in the N-NBI on JT-60U was also extended from the previous ~20 s to 27 s by optimizing the beam steering angle to reduce the grid power loading, as explained in 2.4.2. The time response of the neutron flux from the plasma indicates constant D0 power without degradation during ~30 s. Since the voltage holding capability of the ion sources was as poor as 290 keV due to insufficient conditioning time, the D0 beam was injected from one ion source, and hence the D0 power was no more than 1 MW. Using two negative ion sources, the long-pulse beam at the higher power is to be injected [2.4-1].

2.4.2 R&D Programs of the Negative-Ion-based NBI System for the Performance Improvement

I.2.4-1 There remain three major issues to improve the performance of the negative-ion-based NBI system. One is to reduce the grid power loading below an acceptable level. Second is to understand the electron heat load in the beamline. Third is to improve the voltage holding capability of the ion source, where the usable acceleration voltage has been limited to < 400 kV.

The JT-60U negative ion source was originally designed to produce high current beams of 22 A at 500 keV for 10 s through 1080 apertures that are distributed on five segmented grids. The beamlets must be steered to focus the overall beam envelope. It was found that some of the beamlets were mis-deflected and struck on the acceleration grids, resulting in the high grid power loading. To reduce the grid power loading, a thin plate for tuning the steering angle of the beamlets, called as "field shaping plate (FSP)" was newly designed and tested. Figure I.2.4-1 shows the vertical deflection angle of the outmost beamlets as a function of thickness of the FSP. To suppress the interceptions of the ion beam by the grids and the beam-limiters, a 1 mm thick FSP was chosen and installed on the JT-60U negative ion source. The use of the FSP reduces the grid power loading normalized by the drain power from the previous value of 7.5% to 6%, which is acceptable for the full specifications of the D- ion beam [2.4-2].

I.2.4-2 In the JT-60U negative ion source, the electrons are mainly deflected downwards by the stray magnetic field. Some of the electrons are ejected from the ion source and dissipated on the inertial-cooled parts. Therefore, the ejected electron power should be quantified for long pulse operation. To measure the electron power, a thin stainless steel plate was set below the beam path, and the surface temperature of the plate was measured by infra-red camera (Fig. I.2.4-2(a)). Figure I.2.4-2(b) shows the typical temperature profile and the heat flux on the plate when 300 keV and 3.4 A beam was produced from only the central segment in the upper ion source. It is found that the highest heat flux from one segment is < 8 W/cm2. Its total power load is ~2.6% of the drain power. For full beam with five segments, the highest heat flux is estimated to be ~37 W/cm2 by assuming the heat flux distributions for the other segments to be the same as that measured from the central segment. This power loading can be removed readily by inertia cooling even for 30 s long pulse operation [2.4-3].

The D0 beam power is restricted due to poor voltage holding capability of the ion sources. Even after sufficient conditioning, the achieved acceleration voltage is < 400 kV. Therefore, the voltage holding capability should be increased to the rated value of 500 keV. As the first step for the improvement of the voltage holding capability, the breakdown location of the JT-60 negative ion source has been examined. It is found that the breakdown location varied with the conditioning stage. In the early stage, breakdown occurs mainly in vacuum at gaps between the grids and their support frames with the total surface area of 2.5 m2. Careful observation shows that the conditioning gradually progresses with the breakdowns occurring at many different locations over the large surface area and with the reduction of the outgassing from the grids and the frames. This result suggests that the baking of the grids and frames would be effective to shorten the conditioning time. Over ~400 kV after conditioning of several months, the breakdown location is changed to the surface of the FRP insulator with an inner diameter of 1.8 m. It is recently found in Saitama University that the flashover voltage of a low outgassing epoxy resin (~10-4 Pa⋅m/s) is twice higher than that of the conventional one. The use of the low outgassing epoxy resin is expected to improve the voltage holding capability [2.4-4, 5].

2.4-1 Hanada, M., et al., Rev. Sci. Instrum. 79, 02A519 (2008).
2.4-2 Ikeda, Y., et al., "Recent R&D activities of negative ion based ion source for JT-60 SA," to be published in IEEE Transactions on Plasma Science, 2008.
2.4-3 Kamada, M., et al., Rev. Sci. Instrum. 79, 02C114 (2008).
2.4-4 Hanada, M., et al., "Power loading of electrons ejected from the JT-60 negative ion source," to be published in IEEE Transactions on Plasma Science, 2008.
2.4-5 Kobayashi, K., et al., "Conditioning characteristic of DC 500 kV large electro-static accelerator in negative-ion-based NBI on JT-60U," Proceedings of 23rd international symposium on Discharges and Electrical Insulation in Vacuum, Bucharest, 2008.

2.5 Radio-Frequency Heating System

The FY2007 was quite fruitful year for the JT-60U radio-frequency (RF) heating system. Pulse length injected to the plasma by the electron cyclotron heating (ECH) system reached 30 s of the middle term (-2009) objective over the 25 s of the annual objective. Moreover in the ECH system, the world highest power output of 1.5 MW for 1 s was achieved by a gyrotron to the dummy load. Performance of the RF heating system has been constantly improved to extend the parameter region of experiments.

2.5.1 Long-Pulse Operation of the ECH System

I.2.5-1 The extension of the pulse duration of the ECH system has been tried to enhance the plasma performance in the recent experiment campaign in JT-60U focusing on long sustainment of high performance plasmas. Improvements of the vacuum pumping system of the transmission lines has been carried out in order to avoid pressure rise in the transmission lines due to temperature rise of the components. Vacuum pumping unit was installed at the Matching Optics Unit (MOU) individually for each transmission line. By means of the techniques of controlling heater current and anode voltage during the pulse to keep the oscillation condition, pulse duration of 30 s at 0.4 MW (at gyrotron) has been achieved as shown in Fig. I.2.5-1. Temperature increase of the transmission line components shows availability of the present system for 100 s operation in JT-60SA only for one shot. While upgrade of the cooling system of waveguides will be required to repeat 100 s operations.

2.5.2 High Power Test of the Gyrotron

Improvements of the ECH system (1 MW for 5 s per unit) had been performed until FY2006 toward the higher power and the longer pulse. A Si3N4 ceramic DC-break was introduced into a gyrotron instead of an alumina ceramic DC-break due to the higher thermal strength. In order to avoid heating of a bellows which enables to move the last mirror in the gyrotron, a cover was installed around the bellows to reflect stray-RF. Cooling water of a cavity wall was increased by about 20 %. Moreover, a high-power dummy load system (1.5 MW, CW) was developed to measure the oscillation power from the gyrotron. Those improvements enabled the ECH system to try high power oscillation up to 1.5 MW. In July of 2007, 1.5 MW for 1 s oscillation was achieved for the first time by means of fine optimization of the oscillation parameters as shown in Fig. I.2.5-2 [2.5-1]. The pulse length was not limited by any interlock signals. Therefore, it will be possible to obtain a longer pulse by making more fine adjustments.

2.5.3 Improvement and Performance of the LHRF System

The LHRF (Lower Hybrid Range of Frequency) experiments such as real-time control of plasma current profile were performed. In FY 2006, 6-modules out of 8-modules of the launcher had been used because of damage at the launcher mouth due the malfunction of the arc detector in 2005. In the August of 2007, one of the heavily damaged mouths was repaired carefully from inside of the vacuum vessel of JT-60U, and the protection function against the arcing was upgraded with a new infrared camera watching at the launcher mouth from a tangential port. The conditioning operation for the launcher having 7-modules started in September and the injected power was reached ~ 2 MW in March 2009. The improvement of the power from the achievement of 1.6 MW in 2006 was 1.25 which was more than estimation (1.17~7/6) by a recovered module as shown in Fig. I.2.5-3.

Fig.I.2.5-2      Fig.I.2.5-3

2.5-1 Kobayashi, T., et al., Plasma Fusion Res., 3, 014 (2008).

2.6 Diagnostics Systems
2.6.1 Real-Time Measurement and Feedback Control of Ion Temperature Profile and Toroidal Rotation

Toward the steady-state operation with high beta and high thermal confinement, the real-time measurement and feedback control systems have been developed.

A fast charge exchange recombination spectroscopy (CXRS) system has been developed for the real-time measurement and feedback control of ion temperature (Ti) profile and toroidal rotation velocity (Vt) [2.6-1]. In order to control Ti and Vt in real-time, the charge exchange recombination spectroscopy with high time resolution, the real-time processor system, and the real-time control system have been developed (Fig. I. 2.6-1). Utilizing this system, real-time control of the Ti gradient has been demonstrated with NBs at high beta plasmas (βN~1.6-2.8). The strength of the internal transport barrier is controlled (Fig. I. 2.6-2). Moreover, the real-time control of Vt has been demonstrated from counter to direction. Then the behavior of ELM changed by controlling the Vt.

Fig.I.2.6-1      Fig.I.2.6-2

2.6.2 Zeeman Polarimetry using Lithium Beam Probe for Edge Current Measurement

A lithium ion beam probe has been developed for edge current measurement (Fig. I.2.6-3). A lithium ion gun has been designed by the numerical simulation taking the space charge effects into account because a Zeeman polarimetry requires low beam divergence angle. A porous tungsten disk heated by an electron beam is utilized for an ion emitter. The concave surface of the disk and a peaked heating profile of the electron beam are selected to make a beam focusing better [2.6-2]. Performances of the ion gun have been investigated on a test stand [2.6-3]. A beam current of 10 mA and a divergence angle of 0.2 degrees and equivalent current of 3 mA at the observation area are attained by the ion gun. Figure I.2.6-4 shows the obtained current is increased with the extracted current. Moreover, a long pulse operation of 50 seconds with beam current of 10 mA has been demonstrated. Then the ion gun has been installed and operated on JT-60U successfully. After beam conditioning, a first signal of the lithium beam emission (2 2S - 2 2P resonance line) has been obtained in JT-60U plasma. Adjusting etalon filters, the lithium beam probe system is operated for the edge current measurement.

Fig.I.2.6-3      Fig.I.2.6-4

2.6.3 Polarization Interferometer for Thomson Scattering

I.2.6-5 Recently, the use of a polarization interferometer based on Fourier transform spectroscopy has been proposed for Thomson scattering diagnostics [2.6-4]. It is possible that this method alleviates some of the disadvantages of conventional grating spectrometers. Furthermore, this method delivers a simple and compact system. We are developing the polarization interferometer for Thomson scattering diagnostics with YAG laser to demonstrate the proof-of-principle. For thermal electrons, the optical coherence of the Thomson scattered light at an appropriately chosen optical path delay, is a unique function of Te and ne. The detection system utilizes a single bandpass filter combined with imaging optics and dual detectors to simultaneously observe both dark and bright scattered light interference fringes. The normalized intensity difference between the bright and dark interference fringes gives a direct measure of Te. A schematic of the polarization interferometer for Thomson scattering diagnostics is shown in Fig. I.2.6-5. Scattered light is collected and introduced to the polarization interferometer through a fiber-optic bundle. This polarization interferometer consists of an objective lens as the fiber coupling optics, a band pass filter, a polarizer, a birefringent plate which gives optical path delay, a Wollaston prism, an imaging optics to detector, and dual APD (silicon avalanche photodiode) detectors. Since proof-of-principle tests will be carried out in TPE-RX using the existing YAG laser Thomson scattering system before experiments in JT-60U, parameters for design of a prototype polarization interferometer are fixed as follows: Te ≤ 1 keV, ne ≥ 5×1019 m-3, scattering angle 90. In an initial test using a blackbody radiation source, the magnitude of the change in fringe visibility agrees with the numerical calculation. This result confirms that, following suitable calibration, we will be able to sense visibility changes due to changes in the electron temperature.

2.6-1 Yoshida, M., et al., "Real-time measurement and feedback control of ion temperature profile and toroidal rotation using fast CXRS system in JT-60U," submitted to Fusion Eng. Des.
2.6-2 Kojima, A., et al., Plasma Fusion Res., 2, S1104 (2007).
2.6-3 Kojima, A., et al., "Development of a High-Brightness and Low-Divergence Lithium Neutral Beam for a Zeeman Polarimetry on JT-60U," to be published in Rev. Sci. Instrum.
2.6-4 Hatae, T., et al., Plasma Fusion Res., 2, S1026 (2007).

2.7 Safety Assessment
2.7.1 Application Works for Operational Modifications on JT-60

Based on the law concerning prevention from radiation hazards due to radioisotopes, etc., licensing procedures have been done regarding two items below.

(1) Surface Analyses of Ferritic Steel First Wall Tiles
In FY2005, 1122 ferritic steel (8Cr-2W-0.2V) tiles were installed at the out board first wall inside the vacuum vessel in order to reduce toroidal magnetic field (TF) ripples. For the surface analysis of these tiles removed from the vessel, an application for use of radioisotopes induced on the tiles has been filed to the radiation regulation division of Ministry of Education, Culture, Sports, Science and Technology (MEXT) in March 2007 and was subsequently permitted in April.

(2) Extension of Neutral Beam Injection Time
Pulse lengths of neutral beam injectors (NBI) were extended from 30 s to 60 s in order to meet experiments with long pulse plasma discharges up to 60 s. An application for the NBI pulse extension was permitted in June 2007 and the experiments started in September elongating the beam pulses.

2.7.2 Radiation Safety Assessment for JT-60 Decommissioning

(1) Radiational Exposure during Decommissioning
Potential radiational exposure to workers in the torus hall during disassembly of the JT-60 was preliminarily assessed for internal and external radiation doses together with procedures of the disassembly. It was found that uses of the method of bubble lubricant during diamond wire sawing and of some Green Room facilities for cutting the main structures including the vacuum vessel sectors shall limit spread of contamination and eliminate pollutant in the torus hall, greatly contributing to minimize their internal exposure. In external radiation exposure, a contact dose level was assessed less than 20 μSv/h inside the vacuum vessel at one year cooling time after a shutdown of JT-60, indicating an acceptable radiation environment for the workers. The amount of internal and external exposure doses for the workers was expected to be 100 times less than that of the regal limit, 1mSv/week inside the controlled area.

(2) Storage and Management for Radio-Activated Materials Removed from Controlled Areas
From both points of view of resources saving and reusable nature of the materials, the radio-activated devices or structures replaced with new equipment are expected to be stored appropriately until the potential clearance rule is put into operation. It was found that some additional controlled on-site facilities or areas other than the existing JT-60 storage building are required for their storage and management.

(3) Estimation of Low Level Waste by a Regulatory Clearance.
I.2.7-1 Low level waste of the JT-60U has been estimated by a regulatory clearance [2.7-1]. The JT-60U consists of main devices including a vacuum vessel, magnetic coils, heating devices such as neutral beam injectors and radio frequency systems. Those structural materials include copper, stainless steel, carbon steel, high manganese steel, inconel 625, ferritic steel, lead and others. The gross weight of the devices is about 6,400 tons.
Neutron and gamma-ray fluxes during an operation were calculated with the ANISN code. Induced activity of the materials was calculated by the ACT-4 code in the THIDA-2 code system at various times after an operational shutdown. The total neutron yields was assumed to be 1.14×1020 n for fourteen operation years of the JT-60U.
Figure I.2.7-1 shows time evolutions of the volume of activated materials after the shutdown. The report of IAEA RS-G-1.7 [2.7-2] was referred to compare clearance levels to the results of the activated nuclides induced on the structural materials. The amount of the low level waste of which activation levels exceed the clearance levels is about 5000 tons just after the shutdown. On the other hand, the amount below the clearance levels is 1400 tons.
The former decreases with time and vanishes at 45 years after the shutdown. Asymmetrically, the latter increases year by year as shown in the figure. In JT-60U, stainless steel SS316 of about 50 tons, of which Cobalt content is 0.2 wt%, are used for the base plates of the first wall inside the vacuum vessel. The major source of the activated level of the waste, therefore, takes about 45 years until less than the clearance level due to the long half-time nuclide 60Co.

2.7.3 Nuclear Shielding Assessment using ATTILA Code

Neutron fluxes of a tokamak fusion device with a cryostat were calculated with the three dimensional (3D) nuclear shielding code, ATTILA. ATTILA is a numerical modeling and simulation code designed to solve the 3D multi-group Sn transport equations for neutrons, charged particles, and infrared radiation on an unstructured tetrahedral mesh. It uses a traditional Sn source iteration technique for solving the first order form for the transport equation.

Figure I.2.7-2 depicts a demonstration result of the total neutron flux distribution for a typical superconducting tokamak with a cryostat. From the results, ATTILA has proved to be useful to analyze the nuclear shielding properties especially for the port or duct streaming of an experimental building.

2.7.4 Development of High Heat Resistant Neutron Shielding Resin

In fusion tokamak devices, temperature near a vacuum vessel is expected to rise up to ~300 C, because the wall conditioning by vessel baking is of crucial importance for plasma discharge operations.

Shielding materials such as polyethylene and concrete are widely used. Polyethylene is the most popular resin for neutron shielding, but the heatproof temperature is fairly low. Concrete is not suitable for the additional shielding material in the restricted space around the center of the devices, while it endure a high temperature environment. The heatproof neutron shielding resins such as KRAFTON-HB4 [2.7-3] and EPONITE [2.7-4] had been developed by 2004. KRAFTON-HB4 was the epoxy-based resin that contains boron to reduce the production of secondary gamma rays. It was developed for future FBR shielding materials. KRAFTON-HB [2.7-5] was improved one with the aim of suppressing the nuclear heating of superconducting coils of the DD nuclear fusion devices. The maximum heatproof temperature of it was 150 C. EPONITE was the Colemanite and epoxy-based resin that contains boric acid. It was developed for a medical equipment PET cyclotron with a heatproof temperature of 200 C.

Based on the previously demonstrated results above, we started the development of new resins against much higher temperature range up to 300 C, and successfully produced the new boron-loaded and heat-resisting resin in 2005 [2.7-6]. The heatproof temperature has been improved by an appropriate mixing of stiffening materials with the epoxy-based resin. In 2007, the research and development of the Gel-type heat-resisting resin were initiated in response to the need for more flexible and light shielding materials expected to be useful in situations where an additional shielding is required in narrow or hard-to-reach areas such as locations of diagnostic collimators [2.7-7]. The results obtained to date have been limited to that of a solid-state sample. The glass transition temperature, an indicator of the heatproof temperature, in the specimen was obtained up to 320 C. A test sample piece of the new resin before gel formation is shown in Fig. I.2.7-3.

Fig.I.2.7-2      Fig.I.2.7-3

2.7-1 Sukegawa, A. M., et al., "Estimation of Low Level Waste by a Regulatory Clearance in JT-60U Fusion Device," 15th International Conference on Nuclear Engineering (ICONE-15) , April 22-26, 2007, Nagoya, JAPAN.
2.7-2 IAEA, SAFETY GUIDE, No. RS-G-1.7.
2.7-3 Ueki, K., et al., Nucl. Sci. and Eng., 124, 455 (1996).
2.7-4 Okuno, K., Radiation Protection Dosimetry, 115, 1-4, 258-261 (2005).
2.7-5 Morioka, A., et al., J. Nucl. Sci. Technol., Supplement 4, 109-112 (2004).
2.7-6 Morioka, A., et al., J. Nucl. Mater., 367-370, 1085 (2007).
2.7-7 Sukegawa, A. M., et al., "High Heat Resistant Neutron Shielding Resin, "11th International Conference on Radiation Shielding (ICRS-11) , April 13-18, 2008 , Callaway Gardens, Pine Mountain, Georgia, US.

3. Domestic and International Collaborations
3.1 Domestic Collaboration

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 National Institute for Fusion Science (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 number of collaborators had increased significantly since FY 2003, and it is kept around 150 for recent five yeas as shown in Fig. I.3.1-1.

In FY2007, 149 persons participated, who came from 19 research organs in Japan. Two leaders, one from university or NIFS and the other from JAEA, occupy each subtheme. The number of research subjects of the joint research was 27 in total in FY 2007, categories of which are shown in Table I.3.1-1. Twenty-two journal papers and 1 paper in conference proceedings were published as a result of the joint research in FY 2007.

Fig.I.3.1-1      Table.I.3.1-1

3.2 International Collaboration

I.3.2-1 Status of collaborative research based on the IEA Implementing Agreement on cooperation on the Large Tokamak Facilities is described first. The result of personnel exchanges among Japan, US and Europe are as follows. The number of personnel exchanges, to which JAEA relates, is 11 in total. There are two personnel exchanges from JAEA to EU, 3 from EU to JAEA, 0 from JAEA to US, and 10 from US to JAEA. JT-60U contributes to ITPA/IEA inter-machine experiments. There are 3 to 7 experiments for each Topical Groups. In 2007, 4 related papers were published in journals and 10 related presentations are accepted for IAEA Fusion Energy Conference in 2008.

As for remote collaboration, JAEA provides JT-60 data within the scope of the proposal document sheet (PDS). There are 8 active PDS (3 with EU, 2 with US, 2 with AUG, 1 with EAST).

Remote experimental system (RES) with high network security has been developed in JT-60U. The remote experimental system is produced by personal authentication with a digital certificate and encryption of communication data to protect the JT-60U supervisory control system against illegal access. Remote experiment in JT-60U was demonstrated from Kyoto University (Japan) in 2006 and internationally from IPP Garching (Germany) in 2007 on the occasion of an neo-classical tearing mode stabilisation similarity experiment between JT-60U and ASDEX Upgrade. RES was successfully verified on its authentication, encryption and turn around time. The gateway server blocked all access except the IT-Based Laboratory InfraStructure (ITBL-IS) and Atomic Energy Grid Infrastructure (AEGIS) client certificate, and we confirmed that its authentication mechanism was working properly. The use of packet capture software proved that packets were encrypted. The amount of communication data to display a discharge condition reference page was measured with packet capture software. Throughput was calculated from the window size and the measured Round Trip Time (RTT), and the turn around time was measured from the amount of communication data and throughput. When RTT was 290 milliseconds, the turn around time was 4.13 seconds for the remote experiment from IPP Garching. This gave the applicable response to the remote participants, and it provided mostly the same environment as the onsite researcher. Results are great advances towards the remote experiment in ITER.

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