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The effects of the wave accessibility condition on lower hybrid current drive (LHCD) are investigated in JT-60U. The density dependence of the detectable maximum photon energy of hard X rays in LHCD plasmas is consistent with that of the maximum electron energy which is limited by the accessibility condition. The current drive efficiency and the driven current profile are affected by the accessibility condition. Inaccessible wave power enhances impurity content, particle recycling and main plasma radiation loss, and in extreme cases triggers a MARFE. The wave intensity on the plasma periphery, which is detected by a Langmuir probe located on the divertor plate, is also consistent with an inaccessible power fraction going into the plasma interior. A ray trace analysis taking account of toroidal effect indicates that the inaccessible wave power is localized on the plasma periphery and finally propagates into the divertor plasma region

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

This paper summarizes the main achievements of the RFX fusion science program in the period between the 2008 and 2010 IAEA Fusion Energy Conferences. RFX-mod is the largest reversed field pinch in the world, equipped with a system of 192 coils for active control of MHD stability. The discovery and understanding of helical states with electron internal transport barriers and core electron temperature >1.5 keV significantly advances the perspectives of the configuration. Optimized experiments with plasma current up to 1.8 MA have been realized, confirming positive scaling. The first evidence of edge transport barriers is presented. Progress has been made also in the control of first-wall properties and of density profiles, with initial first-wall lithization experiments. Micro-turbulence mechanisms such as ion temperature gradient and micro-tearing are discussed in the framework of understanding gradient-driven transport in low magnetic chaos helical regimes. Both tearing mode and resistive wall mode active control have been optimized and experimental data have been used to benchmark numerical codes. The RFX programme also provides important results for the fusion community and in particular for tokamaks and stellarators on feedback control of MHD stability and on three-dimensional physics. On the latter topic, the result of the application of stellarator codes to describe three-dimensional reversed field pinch physics will be presented. © 2011 IAEA, Vienna.

In the WT-3 tokamak, the toroidal plasma current is started and ramped up to 6.3 kA by the electron cyclotron (EC) wave alone, without ohmic heating (OH) power. After generation of the plasma current by the EC wave, the OH current can be started up with a very low ohmically induced electric field, i.e.

アクリル酸およびエチレンジアクリレートのプラズマ開始重合によりポリアクリル酸を合成した。PAにより希土類元素イオン (Eu(III), Ce(III), Y(III), Ho(III)およびEr(III)) の吸, 脱着性について検討した。バッチ法による吸着実験でPAが上記の希土類元素吸着性を有し, PAへの希土類元素の吸着はpHに依存することが示された。PA上の希土類元素は0.1M-EDTA (pH 8.5) により定量的に脱着された。PA上への希土類元素の吸着およびPAからの脱着速度はカラム法へPAを応用するのに十分な速さであった。希土類元素の吸, 脱着はPAを充てんしたカラムによって定量的になされた。Eu(III)およびCe(III)の吸着に及ぼす共存陰イオン(Cl-, Br-, SO32-, SO42-, S2O32-, SCN-, ClO4-, HPO42-, H2PO4-, CH3COO-) の影響について検討した。その結果, Eu(III) あるいはCe(III)の吸着はその100倍モルの上記陰イオンが共存しても影響されなかった。

リパーゼによるエステル合成に関しては種々の方法が検討されている.すなわち, 有機溶媒に可溶な修飾酵素を用いる方法, 逆相ミセルに酵素を閉じ込め, 有機相中にマイクロエマルジョンとして分散させる方法, シリカ等の担体に酵素を固定化させる方法等である.この内, 固定化酵素による方法は酵素の安定性に優れることで注目されている.すでに筆者らは, アクリルアミド (以下AAMと略記) とN, N'-メチレンビスアクリルアミド (以下MBAAと略記) のプラズマ開始重合によって得られる超高子量のポリアクリルアミドゲル内にグルコアミラーゼが効率良く固定される事実を見い出している.本研究においては, プラズマ開始重合による固定化法をリパーゼに適用し, 固定化率, 安定性, エステル合成化率について検討した.用いた反応系は医薬, 食品等の乳化剤として用途の広いモノグリセリドを酵素的に合成する目的でグリセリンとオレイン酸からなる系を選んだ.

Direct loss of energetic electrons during lower hybrid current drive in JT-60U is identified through measurements of X-ray signals from the divertor plates. It is found that the lost power decreases with increasing electron density and with increasing launched wave N|| and is characterized by the slowing down time of the electrons accelerated by lower hybrid waves. At ne ~ 0.5 × 1019 m-3 and N|| ~ 1.4, about 5-10% of the injected power is lost through the energetic electrons; the maximum current drive efficiency is obtained at a plasma current of 1.2 MA in JT-60U. It is also found that the confinement of high energy electrons is improved by increasing the plasma current. These results suggest that the power lost directly through energetic electrons will be negligible in ITER in steady state operation

To understand the physics of lower hybrid current drive (LHCD), an investigation into the wave spectrum gap was carried out in JT-60, using two lower hybrid wave launchers. From one launcher a wave with high phase velocity was injected; this wave can interact with high energy electrons to drive the current. Simultaneously, from the other launcher a wave with slower phase velocity was injected; this wave can interact with the bulk electrons. A high energy neutral beam was also injected in order to investigate the interaction of lower hybrid waves with ions. The results show that a fraction of the high phase velocity wave which is upshifted to fill the spectrum gap in the case where only one wave is injected is replaced by a slower wave when two waves are injected. As a result, the tail electron parameters (tail electron density and/or average tail electron energy) were improved while the interaction of the waves with ions decreased, because the frequency of the slower wave was chosen so that the wave did not interact with ions

プラズマ開始重合法によりポリアクリルアミド樹脂 (PAAM) を合成し, 金属イオン (銅 (II), 亜鉛 (II), コバルト(II), ニッケル (II), カドミウム (II), パラジウム (II), 鉄 (III)) の吸, 脱着および濃縮について検討した。PAAMは上記金属イオンをほぼ定量的に吸着し, また銅 (II), 亜鉛 (II), コバルト (II), カドミウム (II), ニッケル (II) の吸着速度は速く, 数分以内に吸着平衡に達することを知った。さらに, PAAMを充てんしたカラムを調製してこれら吸着速度の速い金属イオンの吸, 脱着を検討し, 溶離剤として1M塩酸を用いることによって定量的な回収結果を得, 約10倍の濃縮率を得た。また, 銅 (II) の吸着に及ぼす陰イオンの影響についても検討した。

Monodispersed polymer spheres having irregular surface were prepared using monodispersed polystyrene spheres of 10μm diameter as nucleus. They were then patched with n-butyl methacrylate polymer particles by dispersion polymerization of n-butyl methacrylate.The surface condition of the monodispersed polymer spheres was controlled by controlling the mixing rate during the dispersion polymerization.More uniformly patched and more minute polymer particles were recognized on the surface of each sphere at higher mixing rates.

The role of electron cyclotron resonance heating (ECRH) on the toroidal rotation velocity profile has been investigated in the JT-60U tokamak device by separating the effects of the change in momentum transport, the intrinsic rotation by pressure gradient, and the intrinsic rotation by ECRH. It is found that ECRH increases the toroidal momentum diffusivity and the convection velocity. It is also found that ECRH drives the codirection (co) intrinsic rotation inside the EC deposition radius and the counterdirection (ctr) intrinsic rotation outside the EC deposition radius. This ctr rotation starts from the EC deposition radius and propagates to the edge region.

A plasma current of up to 70 kA has been sustained in WT-3 discharges by second harmonic (2Ωe) electron cyclotron (EC) waves alone, with zero loop voltage, after shutting off the Ohmic heating power. Further, in the case of high power EC wave injection, ramp-up discharges have been obtained. Pulse height analysis of hard X-rays in the line of sight at various angles to the toroidal field shows that the velocity distribution function of the high energy tail electrons is asymmetric in the toroidal direction. The 2Ωe EC wave is mainly absorbed by the tail electrons, and a 2Ωe EC driven current is generated by enhancing the asymmetry of the distribution. The figure of merit of 2Ωe EC current drive (ECCD) is ηEC(2) = (3.2-6.4) × 10−2 (1019 A/Wm2), which is one order of magnitude smaller than that of lower hybrid current drive in WT-3. This low value of ηEC(2) can be attributed to low confinement of the current carrying, high energy tail electrons produced by 2Ωe ECCD.

Confinement properties of lower hybrid current driven (LHCD) plasmas and electron cyclotron heated (ECH) plasmas in the WT-3 tokamak are studied. The outward particle and energy fluxes are estimated from parameters of the scrape-off layer (SOL) plasma. During the LHCD period the density and temperature of the SOL plasma decrease drastically, while the density of the core plasma increases, suggesting a decrease of the outward flux. These results show that particle confinement is improved by LHCD. On the contrary, during ECH the density of the SOL plasma increases and the density of the core plasma decreases, suggesting a deterioration of particle confinement. Improved confinement is not obtained near the density limit of LHCD; the improvement is not related with MHD activities during LHCD.

Sawtooth oscillations (STO) in the Ohmically heated (OH) WT‐3 tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1988 (IAEA, Vienna, 1989), Vol. 1, p. 563] are strongly modified or suppressed by localized electron cyclotron resonance heating (ECH) near the q=1 surface, where q refers to the safety factor. The effect of ECH is much stronger when it is applied on the high‐field side, as compared to the low‐field side. Complete suppression of the STO is achieved for the duration of the ECH when it is applied on the high‐field side, in a low‐density plasma, provided the ECH power exceeds a threshold value. This threshold decreases with the increasing safety factor at the limiter qL and the complete stabilization of STO can be obtained even when ECH is applied on the low‐field side in the high‐qL region. It is noticeable that these complete stabilizations can be obtained only by the localized ECH with the elliptic antenna. The STO stabilization is attributed to a modification of the current density...

Lower hybrid waves propagating both parallel and anti-parallel to the toroidal field (bidirectional waves) produce suprathermal electron tails in both directions. The contribution of these tails is additive for the pressure but subtractive for the plasma current Ip. Therefore, the poloidal beta due to tail electrons, βptai1, becomes much larger than that in the usual case of unidirectional current drive, where βptai1 exceeds unity only when Ip is smaller than the Alfven critical current IA. In the WT-3 tokamak, lower hybrid current drive plasmas produced by bidirectional and unidirectional injection have been compared in terms of βp. The experimental results show that a high βp plasma is generated with Ip larger than IA in the bidirectional case.

The distribution function f() of fast electrons produced by lower hybrid current drive (LHCD) is investigated in the WT-3 tokamak, using a combination of measurements of the hard X-ray (HXR) angular distribution with respect to the toroidal magnetic field and observations of the HXR radial profile. The data obtained indicate the formation of a plateau-like region in f() which corresponds to a region of resonant interaction between the lower hybrid (LH) wave and the electrons. The energy of the fast electrons in the peripheral plasma region is observed to be higher than that in the central plasma region under operational conditions with a high plasma current (Ip ≥ 80 kA). At low current (Ip 50 kA), however, the energy of fast electrons is constant along the plasma radius. In the current ramp-up phase, fast electrons are generated in the directions normal to and opposite to the LH wave propagation. The latter case is ascribed to a negatively biased toroidal electric field induced by the current ramp-up. To study the characteristic change of f() for various current drive mechanisms, HXR measurements are performed in electron cyclotron current driven (ECCD) plasma and in Ohmic heating (OH) plasma. In ECCD plasma, the perpendicular energy of fast electrons increases, which indicates that fast electrons are accelerated perpendicularly by electron cyclotron heating. In both LHCD and ECCD plasmas, fast electrons flow in the direction opposite to the wave propagation, while no such fast electrons are formed in OH plasma.

スチレン-ジビニルベンゼン共重合樹脂を1-ニトロソ-2-ナフトール (NSN) のクロロベンゼン溶液に浸し (漬) することによりゲル粒子を調製した。この粒子を充てんしたゲルカラムにより, 純水および河川水に添加した6種類の金属イオンの吸着, 分離, 濃縮について検討した。銅 (II), 亜鉛 (II), カドミウム (II) およびニッケル (II) については適当なpH条件を選ぶことによっていずれもカラムに捕そくされ, 1M塩酸によって定量的に脱着, 回収されて8～50倍に濃縮することができた。ゲル粒子への吸着pH差あるいは金属-NSN錯体の安定度の差を利用して, 他金属イオンと共存する銅 (II) を選択的に分離することができた。

JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies.

Soft X-rays (SXRs) emitted from plasma in the WT-3 tokamak were measured by three SXR detector arrays, viewing the plasma from three different angles. From these experimental data, the authors reconstructed the two dimensional (2-D) contour of constant SXR emissivity as a function of time, using the technique of computer tomography (CT). Computer simulation results show that data from three or more sets of detector arrays are required to reconstruct the 2-D emissivity structure with sufficient accuracy to distinguish Kadomtsev's model from the quasi-interchange (Wesson's) model of sawtooth crash. Further, it is shown that if the noise amplitude is more than 1.5%, reconstruction of the real 2-D SXR emissivity structure is difficult.

In WT-3, lower hybrid waves are injected into a microwave discharge at the electron cyclotron resonance frequency to initiate the toroidal plasma current IP and to sustain it by radiofrequency (RF) power alone without Ohmic heating power (this is called the 'RF tokamak'). First, fast electrons in the range of several kilo-electronvolts are generated by electron cyclotron and lower hybrid power which initiate a weak toroidal plasma current. Then, fast electrons above 100 keV are produced by the lower hybrid waves, and IP increases rapidly (ΔIp/Δt ~ 780 kA·s−1) to almost 13 kA. Subsequently the toroidal plasma current is ramped up slowly (ΔIp/Δt ~ 180 kA·s−1) to > 25 kA for low densities (e < 6 × 1018 m−3). The fraction ∈R (≤8%) of injected lower hybrid power is converted into poloidal magnetic field energy. For high densities (e = (6−12) × 1018 m−3), a quasi-steady-state discharge is obtained with IP = 7−20 kA, and the figure of merit ηCD is about 0.05 × 1019 A·m−2·W−1. The experimental values of ∈R and ηCD are explained by the direct loss of fast electrons.

Recent results of control of current profile and instability using radiofrequency wave in JT‐60U and prediction analysis in JT‐60SA are descried. In JT‐60U, control of current profile in high‐beta regime was demonstrated by using a real‐time system, where the motional Stark effect diagnostic and lower hybrid wave were used as a detector and actuator, respectively. The minimum value of the safety factor was raised from 1.3 to 1.7 so as to follow the commanded value. Complete stabilization of a neoclassical tearing mode (NTM) with the poloidal mode number m = 2 and the toroidal mode number n = 1 was demonstrated using electron cyclotron (EC) current drive. By scanning the location of EC current drive in detail, strong stabilization effect was found for misalignment less than about half of the full island width. In addition, destabilization of the 2/1 NTM was observed for misalignment comparable to the full island width. Simulation of NTM stabilization in JT‐60SA was performed by using the TOPICS code combined with the modified Rutherford equation. The TOPICS simulation showed that complete stabilization can be achieved more effectively by optimizing the EC wave injection angle and modulating the EC wave.Recent results of control of current profile and instability using radiofrequency wave in JT‐60U and prediction analysis in JT‐60SA are descried. In JT‐60U, control of current profile in high‐beta regime was demonstrated by using a real‐time system, where the motional Stark effect diagnostic and lower hybrid wave were used as a detector and actuator, respectively. The minimum value of the safety factor was raised from 1.3 to 1.7 so as to follow the commanded value. Complete stabilization of a neoclassical tearing mode (NTM) with the poloidal mode number m = 2 and the toroidal mode number n = 1 was demonstrated using electron cyclotron (EC) current drive. By scanning the location of EC current drive in detail, strong stabilization effect was found for misalignment less than about half of the full island width. In addition, destabilization of the 2/1 NTM was observed for misalignment comparable to the full island width. Simulation of NTM stabilization in JT‐60SA was performed by using the TOPICS code combin...

Integrated tokamak modelling that enables the simulation of an entire discharge period is indispensable for designing advanced tokamak plasmas. For this purpose, we extend the integrated code TOPICS to make it more suitable for transient analyses in the fast-ion part. The fast-ion Fokker–Planck solver is integrated into TOPICS at the same level as the bulk transport solver so that the time evolutions of the fast ion and the bulk plasma are consistent with each other as well as with the equilibrium magnetic field. The fast-ion solver simultaneously handles neutral beam-injected ions and alpha particles. Parallelisation of the fast-ion solver in addition to its computational lightness owing to a dimensional reduction in the phase space enables transient analyses for long periods in the order of tens of seconds. The fast-ion Fokker–Planck calculation is compared and confirmed to be in good agreement with an orbit following a Monte Carlo calculation. The integrated code is applied to ramp-up simulations for JT-60SA and ITER to confirm its capability and effectiveness in transient analyses. In the integrated simulations, the coupled evolution of the fast ions, plasma profiles, and equilibrium magnetic fields are presented. In addition, the electric acceleration effect on fast ions is shown and discussed.

The integrated simulation framework for toroidal momentum transport is developed, which self-consistently calculates the neoclassical toroidal viscosity (NTV), the radial electric field and the resultant toroidal rotation together with the scrape-off-layer (SOL) physics-based boundary model. The coupling of three codes, the 1.5D transport code TOPICS, the three-dimensional (3D) equilibrium code VMEC and the 3D drift-kinetic equation solver FORTEC-3D, makes it possible to calculate the NTV due to the non-axisymmetric perturbed magnetic field caused by toroidal field coils. Analyses reveal that the NTV significantly influences in JT-60U and holds the key to determine the NTV profile. The sensitivity of the profile to the boundary rotation necessitates a boundary condition modelling for toroidal momentum. Owing to the high-resolution measurement system in JT-60U, the gradient is found to be virtually zero at the separatrix regardless of toroidal rotation velocities. Focusing on , the boundary model of toroidal momentum is developed in conjunction with the SOL/divertor plasma code D5PM. This modelling realizes self-consistent predictive simulations for operation scenario development in ITER.

Current ramp-up with reduced central solenoid (CS) flux consumption in JT-60SA has been investigated using an integrated modeling code suite (TOPICS) with a turbulent model (CDBM). The plasma current can be ramped-up from 0.6 MA to 2.1 MA with no additional CS flux consumption if the plasma current is overdriven by neutral-beam-driven and bootstrap current. A time duration required for the current ramp-up without CS flux consumption becomes as long as 150 s in the scenario we have examined. In order to achieve a current overdrive condition from 0.6 MA, the current drive by a lower energy neutral beam (85 keV) is effective. A higher energy neutral beam (500 keV) cannot be used in this early phase with a low central electron density (~2 × 1019 m−3) due to large shine through loss, while it can be effectively used in the later phase. Therefore, the main current driver should be switched from the lower energy neutral beam to the higher energy neutral beam during the current ramp-up phase. As a result of an intensive auxiliary heating, plasma beta (the ratio of the plasma pressure to the magnetic pressure) becomes high. Ideal MHD instabilities of such high beta plasmas have been investigated using a linear ideal MHD stability analysis code (MARG2D). External kink modes which might affect the core plasma can be stabilized during the current ramp-up if there is a perfectly conducting wall at the location of the stabilizing plate and the vacuum vessel of JT-60SA and the plasma has a broader pressure profile with the H-mode pedestal and the internal transport barrier.

The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ∼ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ∼ 1.0 only for βN = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to ne/nGW ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.

It is found that no current is driven in a central region of a tokamak plasma once the central current density becomes nearly zero ("current hole"), in spite of high electric conductivity, at the current drive by a toroidal electric field and a radio-frequency wave in experiments on the JT-60U tokamak. This is a new, stiff, self-organized structure of a magnetic field in an axisymmetric toroidal plasma.

A low power polychromatic beam of microwaves is used to diagnose the behavior of turbulent fluctuations in the core of the JT-60U tokamak during the evolution of the internal transport barrier. A continuous reduction in the size of turbulent structures is observed concomitant with the reduction of the density scale length during the evolution of the internal transport barrier. The density correlation length decreases to the order of the ion gyroradius, in contrast with the much longer scale lengths observed earlier in the discharge, while the density fluctuation level remain similar to the level before transport barrier formation.

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

The recent success in coupling lower hybrid (LH) waves in high performance plasmas at JET together with the first demonstration on FTU of the coupling capability of the new passive active multijunction launcher removed major concerns on the possibility of using LH on ITER. LH exhibits the highest experimental current drive (CD) efficiency at low plasma temperature thus making it the natural candidate for off-axis CD on ITER where current profile control will help in maintaining burning performance on a long-time scale. We review recent LH results: long internal transport barrier obtained in JET with current profile sustained and controlled by LH acting under real time feedback together with first LH control of flat q-profile in a hybrid regime with Te ~ Ti. Minutes long fully non-inductive LH driven discharges on Tore Supra (TS). High CD efficiency with electron cyclotron in synergy with LH obtained in FTU and TS opening the possibility of interesting scenarii on ITER for MHD stabilization. Preliminary results of LH modelling for ITER are also reported. A brief overview of ITER LH system is reported together with some indication of new coming LH experiments, in particular KSTAR where CW klystrons at the foreseen ITER frequency of 5 GHz are being developed.

Two topics of applications of RF waves to current profile control in JT‐60U are presented; application of lower‐hybrid (LH) waves to safety factor profile control and electron cyclotron (EC) waves to neo‐classical tearing mode (NTM) control. A real‐time control system of safety factor (q) profile was developed. This system, for the first time, enables 1) real time evaluation of q profile using local magnetic pitch angle measurement by motional Stark effect (MSE) diagnostic and 2) control of current drive (CD) location (ρCD) by controlling the parallel refractive index N∥ of LH waves through control of phase difference (Δφ) of LH waves between multi‐junction launcher modules. The method for real‐time q profile evaluation was newly developed, without time‐consuming reconstruction of equilibrium, so that the method requires less computational time. Safety factor profile by the real‐time calculation agrees well with that by equilibrium reconstruction with MSE. The control system controls ρCD through Δφ in suc...

A new operational scenario of advanced tokamak formation was demonstrated in the JT-60U tokamak. This was accomplished by electron cyclotron and lower hybrid waves, neutral beam injection, and the loop voltage supplied by the vertical field and shaping coils. The Ohmic heating (OH) solenoid was not used but a small inboard coil (part of the shaping coil), providing less than $20%$ of total poloidal flux, was used. The plasma thus obtained had both internal and edge transport barriers, with an energy confinement time of 1.6 times H-mode scaling, a poloidal beta of 3.6, and a normalized beta of 1.6, and a large bootstrap current fraction ($g90%$). This result opens up a possibility to reduce, and eventually eliminate, the OH solenoid from a tokamak reactor, which will greatly improve its economic competitiveness.

Real‐time neoclassical tearing mode (NTM) stabilization system has been developed in JT‐60U. In this system, mode location is first identified by calculating the plasma shape using the Cauchy condition surface method and evaluating electron temperature perturbation profile in real time. Subsequently, steerable mirror angle is changed so that electron cyclotron (EC) wave is absorbed at the mode location. By using this system, a 3/2 NTM has been completely stabilized, and an increase in the beta value and H‐factor has been observed after the stabilization. Effects of EC injection before the NTM onset (‘early EC injection’) have been investigated. It is found that the NTM is more strongly suppressed for the early injection compared with the case of EC injection after the saturation of the NTM growth (‘late EC injection’). It is also found that the suppression effect for the early injection strongly depends on the EC injection angle, as in the case of the late EC injection.

Evolution of the lower-hybrid(LH)-driven current profile was measured during the formation of an internal transport barrier (ITB) in a reversed magnetic shear discharge. As the ITB developed, the initially centrally peaked LH-driven current profile gradually turned hollow and was sometimes accompanied by an off-axis peak in the electron temperature profile. These observations indicate the concentration of LH power deposition to the ITB for this case as a result of nonlinear coupling between the LH waves and the target plasma.

A cooperation framework for analyses and predictions of the neoclassical toroidal viscosity (NTV) and the resultant toroidal flow is developed among the TOPICS, VMEC and FORTEC-3D codes. With the real geometry in JT-60U taken into account, it is found that the NTV is one of the cardinal torque sources especially in the edge region irrespective of the insertion of the ferritic steel tiles (FSTs) that reduce the toroidal field ripple amplitude and is essential to numerically reproduce the measured toroidal rotation profile in the edge. The up–down asymmetric component of the NTV is damped due to the FSTs and the NTV profile correlates with the profile of the radial electric field Er. Predictive simulations for JT-60SA H-mode scenarios are also performed to investigate the effects of the NTV on toroidal rotation. The NTV reversal is observed in the pedestal region where the steep pressure gradient is formed, due to the dependence of the NTV on Er.

One class of internal transport barrier discharges in the JT-60U tokamak is characterized by two relatively flatter regions of the pressure separated by a region with very large pressure gradient. Linear growth rates for toroidal drift-type modes are calculated for discharges in this class, without and with sheared E x B rotation effects. For cases with fully-developed barriers, the results with rotation are consistent with a picture in which the radial electric field generated in part by the steep pressure gradient causes local stabilization, and thus reduction of the local anomalous transport, which allows the steep pressure gradient to persist. If rotation is omitted from the calculation for these cases, or if rotation is included for cases without barriers or with partially developed barriers, the unstable region spreads into the steep pressure gradient region.

ECRF experiments in JT-60U have been started since 1999. ECRF system in JT-60U is characterized by a) collector potential depression (CPD) gyrotrons, b) two diamond windows at the gyrotron output and the vacuum vessel, c) a steerable mirror which enables to change a poloidal injection angle, d) two grooved polarizing mirrors which enable to change elliptical polarization of the incident wave, e) sixty-meter evacuated transmission line with corrugated waveguides. Duration and output power of the gyrotrons have been steadily increased, and ECRF of ∼1.5 MW was successfully injected to the torus for 3 seconds with 3 gyrotrons. By using the ECRF system, the following experimental results are obtained: a) An m/n=3/2 neoclassical tearing mode was completely stabilized by the fundamental ECH/ECCD. The optimum injection angle for the mode suppression was identified by scanning the injection angle during a discharge. Real-time detection of a neoclassical tearing mode and ECRF injection were also performed; b) EC-dr...

A pre-formed negative magnetic shear configuration was sustained stably by lower hybrid current drive (LHCD) for up to 7.5s without MHD instabilities. Location of the minimum safety factor and the internal transport barrier (ITB) have been successfully controlled by chaning the launched wave spectra. Combined current drive experiment with LHCD (2GHz) and the negative-ion based NBCD (0.35 MeV) has shown a significant reduction of LH driven current at the plasma center due to the interaction between LH waves and the injected beam ions. Second harmonic ICRF heating in the negative shear discharges was investigated. The internal transport barrier, which appeared from the pre-NBI heating phase, was maintained for at least 2 seconds by ICRF heating (4–4.5 MW) plus hydrogen NBI heating (∼1.5 MW). The result is in contrast to the NBI heating only case with similar total heating power, where the ITB collapsed within 1 sec. When the ICRF power was comparable to the combined NBI power, a higher stored energy was ach...

Electron density behavior during the fast termination phase of the post-disruption runaway plasma was investigated for the fi rst time. Increases in electron density after the fi rst drop of runaway plasma current were observed. At the second current drop, an increase in electron density with density oscillations having a period of ~1 ms was found. After the second current drop, higher electron density was observed in the case of shorter current decay time.

This paper reports the highlights of the RFX-mod fusion science programme since the last 2010 IAEA Fusion Energy Conference. The RFX-mod fusion science programme focused on two main goals: exploring the fusion potential of the reversed field pinch (RFP) magnetic configuration and contributing to the solution of key science and technology problems in the roadmap to ITER. Active control of several plasma parameters has been a key tool in this endeavour. New upgrades on the system for active control of magnetohydrodynamic (MHD) stability are underway and will be presented in this paper. Unique among the existing fusion devices, RFX-mod has been operated both as an RFP and as a tokamak. The latter operation has allowed the exploration of edge safety factor q edge < 2 with active control of MHD stability and studies concerning basic energy and flow transport mechanisms. Strong interaction has continued with the stellarator community in particular on the physics of helical states and on three-dimensional codes. © 2013 IAEA, Vienna.

Since the first H-mode discharges in 2010, the duration of the H-mode state has been extended and a significantly wider operational window of plasma parameters has been attained. Using a second neutral beam (NB) source and improved tuning of equilibrium configuration with real-time plasma control, a stored energy of Wtot???450?kJ has been achieved with a corresponding energy confinement time of ?E???163?ms. Recent discharges, produced in the fall of 2012, have reached plasma ?N up to 2.9 and surpassed the n?=?1 ideal no-wall stability limit computed for H-mode pressure profiles, which is one of the key threshold parameters defining advanced tokamak operation. Typical H-mode discharges were operated with a plasma current of 600?kA at a toroidal magnetic field BT?=?2?T. L?H transitions were obtained with 0.8?3.0?MW of NB injection power in both single- and double-null configurations, with H-mode durations up to ?15?s at 600?kA of plasma current. The measured power threshold as a function of line-averaged density showed a roll-over with a minimum value of ?0.8?MW at . Several edge-localized mode (ELM) control techniques during H-mode were examined with successful results including resonant magnetic perturbation, supersonic molecular beam injection (SMBI), vertical jogging and electron cyclotron current drive injection into the pedestal region. We observed various ELM responses, i.e. suppression or mitigation, depending on the relative phase of in-vessel control coil currents. In particular, with the 90? phase of the n?=?1 RMP as the most resonant configuration, a complete suppression of type-I ELMs was demonstrated. In addition, fast vertical jogging of the plasma column was also observed to be effective in ELM pace-making. SMBI-mitigated ELMs, a state of mitigated ELMs, were sustained for a few tens of ELM periods. A simple cellular automata (?sand-pile?) model predicted that shallow deposition near the pedestal foot induced small-sized high-frequency ELMs, leading to the mitigation of large ELMs. In addition to the ELM control experiments, various physics topics were explored focusing on ITER-relevant physics issues such as the alteration of toroidal rotation caused by both electron cyclotron resonance heating (ECRH) and externally applied 3D fields, and the observed rotation drop by ECRH in NB-heated plasmas was investigated in terms of either a reversal of the turbulence-driven residual stress due to the transition of ion temperature gradient to trapped electron mode turbulence or neoclassical toroidal viscosity (NTV) torque by the internal kink mode. The suppression of runaway electrons using massive gas injection of deuterium showed that runaway electrons were avoided only below 3?T in KSTAR. Operation in 2013 is expected to routinely exceed the n?=?1 ideal MHD no-wall stability boundary in the long-pulse H-mode (?10?s) by applying real-time shaping control, enabling n?=?1 resistive wall mode active control studies. In addition, intensive works for ELM mitigation, ELM dynamics, toroidal rotation changes by both ECRH and NTV variations, have begun in the present campaign, and will be investigated in more detail with profile measurements of different physical quantities by techniques such as electron cyclotron emission imaging, charge exchange spectroscopy, Thomson scattering and beam emission spectroscopy diagnostics.

The temporal and spatial responses of electron channels (the electron density, ne, and the electron temperature, Te) and ion channels (the ion temperature, Ti, and the toroidal rotation velocity, V) to central electron cyclotron heating (ECH) have been investigated in positive shear H-mode plasmas with a relatively peaked Ti profile and internal transport barrier (ITB) plasmas on JT-60U. Ion temperature decreases with ECH after the increase in the electron temperature in the core region. The time scale of the change in Ti is ≈30–60 ms in H-mode plasmas and almost constant in radius. In ITB plasmas, the time scale is shorter around the ITB foot and becomes longer inside the ITB foot. The experimentally measured causality indicates that the decrease in Ti is consistent with the ion temperature gradient critical gradient reduction. This is also verified through a comparison with linear gyrokinetic stability analyses. The electron heat diffusivity increases with ECH in both H-mode and ITB plasmas, correlating to the increase in the ion heat diffusivity. Electron density with a relatively flat ne profile does not decrease with ECH. On the other hand, the electron density with a peaked ne profile decreases with ECH. The flattening of the ne profile is observed after the increase in the electron temperature in the core region. The time scale of the change in ne is about 200–350 ms. Linear gyrokinetic stability analyses imply that the growth rate of the trapped electron modes, which increase outward particle flux, becomes more pronounced during ECH. The counter intrinsic rotation with ECH is identified on H-mode plasmas with a small torque input (BAL-NBI). The counter intrinsic rotation is generated after the increase in the electron temperature and correlates to the change in the electron temperature with ECH around the EC deposition. The radial region where the counter intrinsic rotation is observed is wider than the radial region where the electron temperature varies with ECH. The time scale of the change in the toroidal rotation velocity is about 90–200 ms around the ECH deposition and longer than the time scale of the change in Te and Ti.

A toroidal momentum solver, which calculates the evolution of the toroidal angular momentum density summed over thermal species, is modelled and integrated into the 1.5D integrated transport code TOPICS. A non-iterative scheme to uniquely determine the radial electric field Er is developed. It is also used to compute the parallel and toroidal flows for each species based on the neoclassical transport theory. The combination of TOPICS and the orbit-following Monte Carlo code OFMC enables us to self-consistently predict the evolution of not only the density, temperature and safety factor but also the toroidal momentum and Er. The framework developed has been tested against JT-60U experiments and showed the predictive capability of toroidal rotation. Several time-dependent simulations in which toroidal rotation and Er play an important role are carried out, showing that the H-mode confinement considerably depends on the direction of toroidal rotation via the change in Er, as observed in JT-60U. Rapid toroidal rotation can be hardly expected in ITER if rotation is solely driven by neutral beam (NB) injection, whereas the residual stress may potentially generate the torque comparable to or greater than that by NBs.

Since its inception in 2002, the International Tokamak Physics Activity topical group on Integrated Operational Scenarios (IOS) has coordinated experimental and modelling activity on the development of advanced inductive scenarios for applications in the ITER tokamak. The physics basis and the prospects for applications in ITER have been advanced significantly during that time, especially with respect to experimental results. The principal findings of this research activity are as follows. Inductive scenarios capable of higher normalized pressure (beta(N)>= 2.4) than the ITER baseline scenario (beta(N) = 1.8) with normalized confinement at or above the standard H-mode scaling are well established under stationary conditions on the four largest diverted tokamaks (AUG, DIII-D, JET, JT-60U), demonstrated in a database of more than 500 plasmas from these tokamaks analysed here. The parameter range where high performance is achieved is broad in q(95) and density normalized to the empirical density limit. MHD modes can play a key role in reaching stationary high performance, but also define the limits to achieved stability and confinement. Projection of performance in ITER from existing experiments uses empirical scalings and theory-based modelling. The status of the experimental validation of both approaches is summarized here. The database shows significant variation in the energy confinement normalized to standard H-mode confinement scalings, indicating the possible influence of additional physics variables absent from the scalings. Tests using the available information on rotation and the ratio of the electron and ion temperatures indicate neither of these variables in isolation can explain the variation in normalized confinement observed. Trends in the normalized confinement with the two dimensionless parameters that vary most from present-day experiments to ITER, gyroradius and collision frequency, are significant. Regression analysis on the multi-tokamak database has been performed, but it appears that the database is not conditioned sufficiently well to yield a new scaling for this type of plasma. Coordinated experiments on size scaling using the dimensionless parameter scaling approach find a weaker scaling with normalized gyroradius than the standard H-mode scaling. Preliminary studies on scaling with collision frequency show a favourable scaling stronger than the standard H-mode scaling. Coordinated modelling activity has resulted in successful benchmarking of modelling codes in the ITER regime. Validation of transport models using these codes on present-day experiments is in progress, but no single model has been shown to capture the variations seen in the experiments. However, projection to ITER using these models is in general agreement with the favourable projections found with the empirical scalings.

A plasma equilibrium control simulator has been developed to simulate the control of plasma position and shape as well as plasma current IP. The simulator consists of an equilibrium calculation component and a controller component. The plasma position, shape, and IP are obtained as a result of an equilibrium calculation under the specified poloidal field coil current. The control simulator enables simulation of the control of the position, shape, and IP using the isoflux technique, and it optimizes the control logic of the coil current in JT-60SA. The plasma equilibrium control is simulated during IP ramp-up. The controllability of the last closed flux surface is also validated during IP flat-top. c