Advanced Tokamak Research Articles

Overview of JT-60U results towards the establishment of advanced tokamak operation

Recent JT-60U experimental results towards the establishment of advanced tokamak (AT) operation are reviewed. We focused on the further expansion of the operational regime of AT plasmas towards higher βN regime with wall stabilization. After the installation of ferritic steel tiles in 2005, the high power heating in a large plasma cross-section in which the wall stabilization is expected has been possible. In 2007, the modification of power supply of NBIs improved the flexibility of the heating profile in long-pulse plasmas. The investigation of key physics issues for the establishment of steady-state AT operation is also in progress using new diagnostics and improved heating systems. In weak magnetic shear plasma, high βN ∼ 3 exceeding the ideal MHD limit without a conducting wall () is sustained for ∼5 s (∼3τR) with RWM stabilization by a toroidal rotation at the q = 2 surface. External current drivers of negative-ion based NB and lower-hybrid waves together with a large bootstrap current fraction (fBS) of 0.5 can sustain the whole plasma current of 0.8 MA for 2 s (1.5τR). In reversed magnetic shear plasma, high βN ∼ 2.7 (βp ∼ 2.3) exceeding with qmin ∼ 2.4 (q95 ∼ 5.3), HH98(y,2) ∼ 1.7 and fBS ∼ 0.9 is obtained with wall stabilization. These plasma parameters almost satisfy the requirement of ITER steady-state scenario. In long-pulse plasmas with positive magnetic shear, a high βNHH98(y,2) of 2.6 with βN ∼ 2.6 and HH98(y,2) ∼ 1 is sustained for 25 s, significantly longer than the current diffusion time (∼14τR) without neoclassical tearing modes (NTMs). A high G-factor, (a major of fusion gain), of 0.54 and a large fBS > 0.43 are suitable for ITER hybrid operation scenario. Based on the plasma for ITER hybrid operation scenario, the high βN of 2.1 with good thermal plasma confinement of HH98(y,2) > 0.85 is sustained for longer than 12 s at and frad > 0.79. Physics studies for the development of AT plasmas, physics studies of H-mode, pedestal and ELM characteristics and physics studies on impurity transport, SOL/divertor plasmas and plasma–wall interactions are also in progress. The active NTM stabilization system using modulated ECCD, which is synchronized to rotating island, has been developed and the efficiency of modulated ECCD in m/n = 2/1 NTM stabilization has been demonstrated. The intrinsic toroidal rotation driven by the ion pressure gradient and by the ECH is confirmed. The dedicated H-mode and pedestal experiments indicate two scalings, H-factor evaluated for the core plasma as and pedestal width scaling of . New fast diagnostics with high spatial and temporal resolutions reveals the different structures of pedestal pressure between co- and counter-rotating plasma, resulting in different ELM sizes determined by the radial penetration depth of the ELM crash. The tungsten accumulation becomes more significant with increasing toroidal rotation in the counter-direction.

Balance of the stored energies sustained by the internal and edge transport barriers and effects of ELMs and L–H transitions in JT-60U

To understand key physics processes determining radial profiles of the kinetic plasma parameters in the advanced tokamak operation scenarios, correlations between the edge transport barrier (ETB) and the internal transport barrier (ITB) have been studied in the JT-60U tokamak device. It has been found that the edge pedestal poloidal beta, βp-ped, increases almost linearly with the total poloidal beta, βp-tot, over a wide range of the plasma current for type I ELMing H-mode plasmas, and this dependence becomes stronger with increasing triangularity. This dependence is not due to the profile stiffness, since the dependence is the same regardless of the existence of ITB. As the stored energy inside the ITB-foot radius (WITB) increases, the total thermal stored energy (Wth) increases and then the pedestal stored energy (Wped) increases. On the other hand, as Wped increases, the ELM penetration expands more inwards and finally reaches the ITB-foot radius. At this situation, the ITB-foot radius cannot move outwards because of the erosion by ELMs. Then the fractions of WITB/Wth and Wped/Wth become almost constant. It has also been found that the type I ELM expels/decreases the edge toroidal momentum larger than the edge ion thermal energy. The ELM penetration for the toroidal rotation tends to be deeper than that for the ion temperature and can exceed the ITB-foot radius. The ELM penetration is deeper for CO-rotating plasmas than CTR rotating plasmas. In both cases, the ELM penetration is deeper in the order of the toroidal rotation (Vt), the ion temperature (Ti) and then the electron temperature (Te). The L–H transition also changes the Vt profile more significantly than the Ti profile. At the L–H transition, the pedestal Vt shifts into the CTR-direction deeply and suddenly without a change in Ti, and then the pedestal Vt grows further together with a growth of the pedestal Ti in a slower timescale. Such changes in Vt by ELMs and L–H transitions may affect degradation/evolution of ITBs.

An Evaluation of Fusion Energy R&D Gaps Using Technology Readiness Levels

The ARIES Team currently is engaged in an effort called the “ARIES Pathways Study”. The goals of this study are to evaluate remaining R&D needs toward practical fusion energy and to identify and evaluate possible “next step” devices to bridge the gap between ITER and an attractive power plant. In order to evaluate our current state of readiness and remaining R&D needs, we adopted a methodology called “Technology Readiness Levels”. We defined a quantitative set of readiness levels that encompass the major technology challenges for fusion energy development, and have applied them to evaluate our current level of advancement and R&D needs for an advanced tokamak power plant concept based on recent ARIES designs. Results of the evaluation and recommendations for future R&D are presented.

Design of the DEMO Fusion Reactor Following ITER.

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Kinetic analysis of the resistive wall modes in the ITER advanced tokamak scenario

It is found that n = 1 resistive wall modes in the ITER advanced scenario can be fully stabilized by modestly low rotation with a rotation frequency (normalized to the Alfvén frequency at the magnetic axis) of about Ω = 0.0075. The existence of this stabilization scheme is proved with the AEGIS-K (Adaptive EiGenfunction Independent Solution-Kinetic) code, which provides a fully kinetic (non-hybrid) and self-consistent (non-perturbative) description of the system. Wave-particle resonances, shear Alfvén continuum damping, trapped particle effect and the parallel electric effects are all taken into account. The rotation frequency for full stabilization is much larger than the diamagnetic drift frequency; therefore, finite Larmor radius effects are negligible. We also find that the rotation stabilization window opens first near the ideal wall limit.

Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program “Broader Approach”. The maximum heat flux to JT-60SA divertor is estimated to ∼15 MW/m 2 for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R&D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R&D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m 2 as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

Progress in physics and control of the resistive wall mode in advanced tokamaks

Self-consistent computations are carried out to study the stability of the resistive wall mode (RWM) in DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] plasmas with slow plasma rotation, using the hybrid kinetic-magnetohydrodynamic code MARS-K [Y. Q. Liu et al., Phys. Plasmas 15, 112503 (2008)]. Based on kinetic resonances between the mode and the thermal particle toroidal precession drifts, the self-consistent modeling predicts less stabilization of the mode compared to perturbative approaches, and with the DIII-D experiments. A simple analytic model is proposed to explain the MARS-K results, which also gives a qualitative interpretation of the recent experimental results observed in JT-60U [S. Takeji et al., Nucl. Fusion 42, 5 (2002)]. Our present analysis does not include the kinetic contribution from hot ions, which may give additional damping on the mode. The effect of particle collision is not included either. Using the CARMA code [R. Albanese et al., IEEE Trans. Magn. 44, 1654 (2008)], a stability and control analysis is performed for the RWM in ITER [R. Aymar et al., Plasma Phys. Controlled Fusion 44, 519 (2002)] steady state advanced plasmas, taking into account the influence of three-dimensional conducting structures.

FAST plasma scenarios and equilibrium configurations

In this paper we present the fusion advanced studies torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of plasma–wall interaction, burning plasma physics, ITER relevant operation problems and steady state scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen in a wide range of parameters from high performance H-mode (toroidal field, BT, up to 8.5 T; plasma current, IP, up to 8 MA) to advanced tokamak (AT) operation (IP = 3 MA) as well as full non-inductive current scenario (IP = 2 MA). The coupled heating power is provided with 30 MW delivered by an ion cyclotron resonance heating system (30–90 MHz), 6 MW by a lower hybrid system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an electron cyclotron resonant heating system (170 GHz − BT = 6 T) for MHD and localized electron heating control and, eventually, with 10 MW by a negative neutral ion beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ion induced as well as turbulence fluctuation spectra, thus addressing the cross-scale couplings issue of micro- to meso-scale physics. The non-inductive scenario at IP = 2 MA is obtained with 60–70% of bootstrap current and the remaining by LHCD. Predictive simulations of the H-mode scenarios have been performed by means of the JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and shape control studies are also presented for the reference scenario.

Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M [H. Kawashima, et al., Demonstration of ripple reduction by ferritic steel board insertion in JFT-2M, Nucl. Fusion, 41 (2001) 257–263] and JT-60U [H. Takenaga, the JT-60 Team, Overview of JT-60U results for development of steady-state advanced Tokamak scenario, Proceedings of the 21st IAEA Fusion Energy Conference, Chengdu, China, 2006]. In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called “ferromagnetic inserts”. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.

Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been calculated from charge exchange measurements of impurity ion temperature, density and rotation velocity profiles, using the force balance equation. The ion temperature and the toroidal and poloidal rotation velocities are compared in plasmas with both reversed and optimized magnetic shear profiles. Poloidal rotation velocity (vθ) in the ITB region is measured to be of the order of a few tens of km s−1, significantly larger than the neoclassical predictions. Increasing levels of the TF ripple are found to decrease the ion temperature gradient in the ITB region, a measure for the quality of the ITB, and the maximum value of vθ is reduced. The poloidal rotation term dominates in the calculations of the total radial electric field (Er), with the largest gradient in Er measured in the radial region coinciding with the ITB.

Collisionless trapped electron and ion temperature gradient modes in an advanced tokamak equilibrium

The linear stability of coupled collisionless trapped electron (TE) and ion temperature gradient (ITG) modes is investigated in an International Thermonuclear Experimental Reactor [C. Gormezano et al., Nucl. Fusion 47, S285 (2007)]-like magnetic field configuration. An advanced fluid model in the electrostatic limit and the ballooning mode formalism are used to derive an eigenvalue equation. The growth rates and real frequencies of the most unstable modes and their eigenfunctions are calculated. The coexistence of unstable ITG and TE modes and their relative strengths with respect to the fraction of trapped electrons, wave vector, and local plasma parameters, such as density gradient, electron/ion temperature gradient, and ion to electron temperature ratio are analyzed. The influence of geometrical factors, such as local/global magnetic shear and magnetic field curvature on both ITG and TE modes is also discussed.

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δ B T = 0.08% up to δ B T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β N values above the ‘no-wall limit’ ( β N ∼ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.

The three RF additional heating systems for FAST

FAST (Fusion Advanced Studies Torus) is a new tokamak machine proposed by the Italian Fusion Association as a Satellite Tokamak for the ITER programme. FAST will operate with deuterium plasmas to avoid the complexity deriving from the use of tritium. Therefore burning plasma conditions, where energy density of fast ions and of charged fusion products is a significant fraction of the total plasma energy density, will be achieved by accelerating plasma ions above the half-MeV range through an Ion Cyclotron Resonance Heating (ICRH) system ( P = 30 MW, f = 60–90 MHz). For long pulse Advanced Tokamak (AT) scenarios, a Lower Hybrid Current Drive (LHCD) system ( P = 6 MW, f = 3.7 GHz) has been envisaged to actively control the current profile, whereas an Electron Cyclotron Resonant Heating (ECRH) system ( P = 4 MW, f = 170 GHz) will provide enough RF power for MHD control.

Active control of edge localized modes with a low n perturbation fields in the JET tokamak

Active control of edge localized modes (ELMs) by using static external magnetic perturbation fields with low toroidal mode number, n, has been demonstrated for both, ITER baseline ( q 95 ∼ 3 ) and high beta advanced tokamak scenarios at the JET tokamak. During the application of the low n field the ELM frequency increased by a factor up to ∼4–5. Reduction in carbon erosion and ELM peak heat fluxes on the divertor target by roughly the same factor as the increase of the ELM frequency has been observed. The frequency of the mitigated ELMs using a low n field is found to increase proportional to the total input heating power. Compensation of the density pump-out effect observed when the external low n field is applied has been achieved by gas fueling in low triangularity plasmas.

Plasma impurity content, gas fueling, and exhaust on DIII-D over extended periods between boronizations

During the 2006 and 2007 DIII-D experimental campaigns the rate of boronization events was reduced significantly from past campaigns with no detrimental effects on discharges, including high-performance hybrid and advanced tokamak discharges. Boronizations were completed early in both campaigns due to preceding entry vents. Over the 3-month duration of each campaign, a database of edge and core impurity emission and fueling/exhaust rates from many hybrid discharges was developed that demonstrated little secular change over 7000 plasma-seconds of operations. After 6000 s, a set of seven sequential hybrid discharges was executed with no between-shot helium glow-discharge cleaning. While small effects on fueling and exhaust are observed, density remains controlled and fusion performance is held constant. These results, obtained with the all graphite wall on DIII-D, are promising for the next generation of superconducting, long-pulse tokamaks, where studies of stationary, high performance will be of great interest.

Self-consistent modeling of impurity seeded JET advanced tokamak scenarios

The influence of neon and nitrogen seeding on the divertor heat load in JET tokamak is studied in terms of the numerical simulations of the whole device by the self-consistent core–edge code COREDIV and results are presented for advanced tokamak scenarios. First, the code has been applied to simulate two JET shots (#69980, #69987) characterized by different radiation levels. With this basis the code is used next to provide guidance for the JET experiments in terms of: effects on the heat load at the target plates for various x-point positions with respect to the target plates, and the effect of higher plasma and separatrix density, as well as effects of pure or mixed Ne/ N 2 -seeding.

Divertor heat load in ITER-like advanced tokamak scenarios on JET

Heat loads on the JET MarkIIHD divertor are measured by IR thermography in high triangularity plasmas with edge conditions of advanced tokamak scenarios (AT). AT scenarios operate at low density ( n e , av < 5 × 10 19 m - 3 at JET) to maximise the non inductive current drive, leading to high target temperature ( 20 < T e < 40 eV ), hence high heat loads. Comparisons with a typical baseline inductive scenario ( n e , av ≃ 10 × 10 19 m - 3 ), show that for similar input power ( ≃ 20 MW ) the outer target heat load is higher by a factor 4 in AT scenario, and is above the limit for the ITER-like wall (ILW). Operating long pulses in AT scenarios in the ILW could then be an issue and heat load reduction using impurity seeding is investigated.

An Assessment of the Penetrations in the First Wall Required for Plasma Measurements for Control of an Advanced Tokamak Plasma Demo

An Assessment of the Penetrations in the First Wall Required for Plasma Measurements for Control of an Advanced Tokamak Plasma Demo

Achievement of 1.9 MW for 300 s during the commissioning of the KSTAR ICRF transmitter

KSTAR (Korea Superconducting Tokamak Advanced Research) is a national tokamak aiming at the high beta operation based on AT (Advanced Tokamak) scenarios in Korea and ICRF (Ion Cyclotron Ranges of Frequency) is one of the essential heating and current drive tools to achieve this goal. The ICRF heating and current drive scenario requires 4 units of 2 MW transmitters with a frequency range from 25 to 60 MHz. The first KSTAR transmitter is a modified FMIT (Fusion Material Irradiation Test) transmitter consisting of four amplifier stages. An amplitude-modulated 1 mW frequency source drives a 500 W solid state wideband amplifier, which in turn drives three tuned triode/tetrode amplifier stages. The tube employed in the final power amplifier is a 4CM2500KG tetrode fabricated by CPI (Communications & Power Industries). After the fabrication of the cavity and power supply was completed in 2004, several failures of the tube during a factory and a site acceptance test occurred before eventually achieving 1.9 MW for 300 s at 33 MHz in 2007. The electrical efficiency of the FPA (Final Power Amplifier) is about 70%. Although this is a very encouraging result for the development of an ICRF transmitter for ITER (International Thermonuclear Experimental Reactor), continued efforts for a reliable operation are required to achieve the final goals of the KSTAR and ITER ICRF system.

The technology and science of steady-state operation in magnetically confined plasmas

The steady-state operation of magnetically confined fusion plasmas is considered as one of the ‘grand challenges’ of future decades, if not the ultimate goal of the research and development activities towards a new source of energy. Reaching such a goal requires the high-level integration of both science and technology aspects of magnetic fusion into self-consistent plasma regimes in fusion-grade devices.On the physics side, the first constraint addresses the magnetic confinement itself which must be made persistent. This means to either rely on intrinsically steady-state configurations, like the stellarator one, or turn the inductively driven tokamak configuration into a fully non-inductive one, through a mix of additional current sources. The low efficiency of the external current drive methods and the necessity to minimize the re-circulating power claim for a current mix strongly weighted by the internal ‘pressure driven’ bootstrap current, itself strongly sensitive to the heat and particle transport properties of the plasma. A virtuous circle may form as the heat and particle transport properties are themselves sensitive to the current profile conditions. Note that several other factors, e.g. plasma rotation profile, magneto-hydro-dynamics activity, also influence the equilibrium state.In the present tokamak devices, several examples of such ‘advanced tokamak’ physics research demonstrate the feasibility of steady-state regimes, though with a number of open questions still under investigation. The modelling activity also progresses quite fast in this domain and supports understanding and extrapolation.This high level of physics sophistication of the plasma scenario however needs to be combined with steady-state technological constraints. The technology constraints for steady-state operation are basically twofold: the specific technologies required to reach the steady-state plasma conditions and the generic technologies linked to the long pulse operation of a fusion device. The first includes specific additional heating and current drive methods (through externally launched radio-frequency waves or energetic atoms), fuelling and pumping methods, dedicated plasma diagnostics as well as software technologies required for mandatory real time control loops, involving such actuators and sensors. The second class of technologies, generic to any magnetic fusion device, includes the superconducting magnet technologies, in order to provide a stationary confinement magnetic field, the actively cooled plasma facing components (PFCs) handling either radiated or convected power fluxes (often in excess of several tens of MW m−2), dedicated diagnostics monitoring the interfaces (e.g. infrared survey of PFCs), etc. The detailed specifications of all elements must comply with a reactor-relevant environment, in terms of operational parameters as well as lifetime.The paper presents a summary of the present status and understanding of the technology and science of steady-state operation in magnetically confined plasmas, as well as the forthcoming work programme dedicated to the vast R&D programme undertaken in this domain, in particular within the European fusion framework.