Advanced Tokamak Research Articles

Design of lower divertor for JT-60SA

Construction has just begun in the JT-60 Super Advanced (JT-60SA) tokamak project, which is part of the Japanese domestic program and the Japan – EU international program “ITER Broader Approach”. All plasma facing components (PFCs) will be actively cooled because they are exposed to high heat flux for long duration. A lower single-null closed divertor with a vertical target (VT) will be installed when experiments begin. A divertor cassette integrated with coolant pipe connections will be used for remote handling maintenance. Each cassette covers a 10° sector in the toroidal direction. Modularized PFCs such as VTs, baffles, and domes will be assembled on divertor cassettes. Static structural analysis shows that the displacement and stress in the divertor module are generally small for determining the dead weight, coolant pressure, and electromagnetic forces (EMFs). However, the support structure of the outer baffle requires improvement because eddy currents generate concentrated overturning EMFs.

Fast ion redistribution and losses in JET advanced tokamak scenario**This paper is an extended version of a contribution presented at the 11th IAEA Technical Committee Meeting on Energetic Particles, Kyiv (Ukraine), 21–23 September 2009.

The influence of Alfvén eigenmodes (AEs) (core-localized toroidal Alfvén eigenmode (TAE) and high-m global TAE) on redistribution and loss of fast ions in the advanced tokamak scenario has been investigated in JET dedicated experiments. Temporal evolution of the spatial structure of AEs present in the plasma during q-profile evolution has been calculated with the MISHKA code. On the other hand, fast ion losses with resolution in energy and pitch angle were measured with a scintillator probe, while the internal redistribution of the fast ions was analysed by measuring the temporal evolution of the radial profiles of γ-ray emission coming from the fast ions colliding with the main impurities, Be and C. Correlating the measurements above, it is possible to assess the effect of AEs on redistribution and loss of fast ions. It is found that the TAE localized in the core of the plasma causes a severe redistribution of the fast ions in the plasma centre, leading to a flattening of the vertical profile of γ-ray emission. In addition, the core-localized TAE also causes enhanced fast ion losses. As the core-localized TAE moves radially outwards, a transition to high-m global TAE occurs. In the transformation phase, a coupled mode with components both in the core and in the edge of the plasma exists. The maximum number of losses is normally measured when this coupled mode with a large radial extent is present in the plasma.

MHD stability analysis for advanced tokamak modes in the KSTAR device

The Korea Superconducting Tokamak Advanced Research (KSTAR) device aims to demonstrating the steady-state operation of high-performance advanced tokamak (AT) modes. In order to meet this research goal it is critical to have a good magnetohydrodynamic (MHD) stability, so that KSTAR adopted a strong plasma shape and a conducting wall close to plasma for such stability. An early calculation during the KSTAR design phase had shown that a target AT mode stable up to β N above 5 can be then obtained. A recent work by Katsuro-Hopkins et al. [O. Katsuro-Hopkins, S.A. Sabbagh, J.M. Bialek, H.K. Park, J.G. Bak, J. Chung, et al., Equilibrium and global MHD stability study of KSTAR high beta plasmas under passive and active mode control, Nucl. Fusion 50(2010) 025019] showed, however, that the maximum β N value can be substantially lower than 5, unlike the earlier result. In this work, we present a more detailed study on the MHD stability limit of the KSTAR target AT mode and try to clarify the discrepancy observed in the previous two works. It is shown that in the reverse-shear plasma the target mode with β N above 5 can be obtained if the pressure profile is relatively peaked, but the maximum β N value is substantially reduced below 5 if the pressure profile becomes broader. This result suggests the importance of a proper control of the pressure profile to get the high-beta AT mode in KSTAR.

First results obtained from the soft x-ray pulse height analyzer on experimental advanced superconducting tokamak

An assembly of soft x-ray pulse height analyzer system, based on silicon drift detector (SDD), has been successfully established on the experimental advanced superconducting tokamak (EAST) to measure the spectrum of soft x-ray emission (E=1-20 keV). The system, including one 15-channel SDD linear array, is installed on EAST horizontal port C. The time-resolved radial profiles of electron temperature and K(alpha) intensities of metallic impurities have been obtained with a spatial resolution of around 7 cm during a single discharge. It was found that the electron temperatures derived from the system are in good agreement with the values from Thomson scattering measurements. The system can also be applied to the measurement of the long pulse discharge for EAST. The diagnostic system is introduced and some typical experimental results obtained from the system are also presented.

Ion temperature gradient driven transport in tokamaks with square shaping

Advanced tokamak schemes which may offer significant improvement to plasma confinement on the usual large aspect ratio Dee-shaped flux surface configuration are of great interest to the fusion community. One possibility is to introduce square shaping to the flux surfaces. The gyrokinetic code GS2 [Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1996)] is used to study linear stability and the resulting nonlinear thermal transport of the ion temperature gradient driven (ITG) mode in tokamak equilibria with square shaping. The maximum linear growth rate of ITG modes is increased by negative squareness (diamond shaping) and reduced by positive values (square shaping). The dependence of thermal transport produced by saturated ITG instabilities on squareness is not as clear. The overall trend follows that of the linear instability, heat and particle fluxes increase with negative squareness and decrease with positive squareness. This is contradictory to recent experimental results [Holcomb et al., Phys. Plasmas 16, 056116 (2009)] which show a reduction in transport with negative squareness. This may be reconciled as a reduction in transport (consistent with the experiment) is observed at small negative values of the squareness parameter.

Innovative signal processing and data analysis methods on JET for control in the perspective of next-step devices

In the last few years, it has been realized that more sophisticated control schemes are necessary to push the boundaries of tokamak operation and the performance of reactor-like machines. In addition, JET needs to operate safely with the new metallic wall and such protection will be needed for ITER. These objectives have motivated the development, benchmark and validation of new signal processing and data analysis methods. Two new approaches for the determination of the magnetic topology in real time have been validated on an extensive database of JET discharges, including advanced tokamak scenarios. Robust methods of confinement regime identification and disruption prediction are a prerequisite for safe, general control schemes. New identifiers have been developed and their success rates exceed 99% in determining whether plasmas are in the L or H mode. A new disruption predictor is being developed and has already provided success rates higher than 90% in realistic real-time conditions. Moreover, the generalization capability of this new predictor has been confirmed by applying it to new experimental campaigns not used for the training. The success rate remains high even more than ten campaigns, or about four years, after the last one used for the training. The deployment of video cameras in real time requires the development of new image processing algorithms, which have already been implemented and validated successfully on JET for the real-time identification of hot spots with a time resolution of tens of milliseconds. A series of new feedback schemes has also been explicitly developed not much to control the plasma but to really improve the physics understanding of some phenomena. Particularly interesting are the simultaneous control of the safety factor and pressure profiles and the real-time tracking of toroidal Alfven eigenmode instabilities. These advanced feedback schemes for physics understanding often require more advanced signal processing techniques like adaptive filtering, which have already been implemented. The paper concludes by discussing the use of these real-time analysis and control developments in next-step machines such as ITER.

Advanced tokamak research with integrated modeling in JT-60 Upgrade

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.

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

A demonstration tokamak is an essential next step toward a magnetic fusion–based reactor. One based on advanced tokamak plasmas is especially appealing because of its relative compactness. However, many plasma measurements will be required to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This technical note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.

Saturated ideal modes in advanced tokamak regimes in MAST

MAST plasmas with a safety factor above unity and a profile with either weakly reversed shear or broad low-shear regions, regularly exhibit long-lived saturated ideal magnetohydrodynamic (MHD) instabilities. The toroidal rotation is flattened in the presence of such perturbations and the fast ion losses are enhanced. These ideal modes, distinguished as such by the notable lack of islands or signs of reconnection, are driven unstable as the safety factor approaches unity. This could be of significance for advanced scenarios, or hybrid scenarios which aim to keep the safety factor just above rational surfaces associated with deleterious resistive MHD instabilities, especially in spherical tokamaks which are more susceptible to such ideal internal modes. The role of rotation, fast ions and ion diamagnetic effects in determining the marginal mode stability is discussed, as well as the role of instabilities with higher toroidal mode numbers as the safety factor evolves to lower values.

Integrated Performance and Critical Issues for Steady-State Operation in JT-60U

This paper reports on the integrated performance achieved in JT-60U toward the steady-state operation foreseen in the ITER and DEMO reactors. Advanced tokamak plasmas with weak shear or reversed shear have been optimized to confront critical issues such as high-beta operation with high confinement, the compatibility of high-density operation with high confinement, and long sustainment with a high non-inductive current drive fraction. As a result, high-integrated performance was achieved in both plasma regimes. For example, high-confinement reversed shear plasmas with a high bootstrap current fraction exceeding the no-wall beta limit have been obtained in the reactor relevant q95∼5.3; high values of βN∼2.7, HH98y2∼1.7, ne/nGW∼0.87, and fBS∼0.9 are simultaneously achieved with a reversed q profile with qmin∼2.3.

Rotational stabilization of resistive wall modes in ITER advanced tokamak scenarios

Rotational stabilization of n=1 resistive wall modes in ITER advanced scenarios [K. Ikeda, Nucl. Fusion 47 (2007)] is investigated, where n is the toroidal mode number. In particular, we present numerical results for the ITER strongly reversed shear case, in comparison to the weakly reversed shear case. The rotation frequency is assumed to be modestly low. Our investigation employs the adaptive eigenfunction independent solution-kinetic (AEGIS-K) code [L. J. Zheng et al., “AEGIS-K code for linear kinetic analysis of toroidally axisymmetric plasma stability,” J. Comput. Phys. (to be published)], which provides a fully kinetic (nonhybrid) and self-consistent (nonperturbative) description. AEGIS-K includes wave-particle resonances, shear Alfvén continuum damping, trapped particle effects, and parallel electric effects, but not finite Larmor radius effects. In the case without rotation and kinetic effects included, we find that the strongly reversed shear configuration is more favorable for perfectly conducting wall stabilization of resistive wall modes, in that it has a higher conducting wall beta limit than the weakly reversed shear case. With sufficient rotation, the strongly reversed shear case can also achieve a higher beta limit for completely suppressing the resistive wall modes. However, the marginal rotation frequency required for complete resistive wall mode stabilization in the strongly reversed shear case is about twice as high as that required for the weakly reversed shear case.

Mission and Overview of a Fusion Development Facility

The objective of the Fusion Development Facility (FDF) under consideration is to fill the gaps between ITER and current experiments and a fusion demonstration power plant (DEMO). FDF should carry forward advanced tokamak (AT) physics and enable development of fusion's energy applications. Near-term AT physics and nonsuperconducting magnet technology will be used to achieve steady state with burn, producing 100-250 MW fusion power with modest energy gain (Q<5) in a modest sized device (between DIII-D and Joint European Torus). FDF will further develop all elements of AT physics for an advanced performance DEMO. With neutron flux at the outboard midplane of 1-2 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , continuous operation for periods of up to two weeks, and a goal of a duty factor of 0.3 per year, FDF can produce fluences of 3-6 MW·year/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> in ten years of operation, for fusion nuclear component research and development. The development of blankets suitable for tritium, electricity, and hydrogen production will be done first in port modules. Then, the most promising candidates will be deployed as full blankets in FDF. Two to three full blankets and about a dozen port-blanket types could be tested. The goal of FDF is to demonstrate closure of the fusion fuel cycle, producing its own tritium. FDF, ITER, International Fusion Materials Irradiation Facility, and other AT devices will provide the basis for a fusion DEMO power plant of the ARIES-AT type.

Implications for ITER CODAC from DIII-D experience

The DIII-D digital plasma control system (D3PCS) has been in use since 1993 controlling DIII-D plasmas. Control research and D3PCS development at DIII-D has revealed aspects of advanced tokamak control which can help inform the ITER design. The D3PCS has also been adapted for use at several fusion devices worldwide, which has allowed the DIII-D team to obtain a significant understanding of common requirements for plasma control on multiple fusion devices, along with substantial experience in the alternative computing, data acquisition, and networking technologies presently available. We describe some of what has been learned and highlight the relevance of these lessons for ITER Control, Data Access and Communication (CODAC). Several capabilities of the D3PCS are described in the context of ITER requirements. This description touches on multiple subsystems described in the CODAC conceptual design. We also discuss features of D3PCS architecture that are appropriate for today's experimental devices but may not be appropriate for ITER, which requires a much more comprehensive system for ensuring device safety. Even for these applications, the knowledge gained in implementing methods to aid in ensuring device safety on present devices provides useful guidance for an eventual ITER solution.

Development status of KSTAR 5 GHz LHCD system

A steady-state lower hybrid current drive (LHCD) system is under development for advanced tokamak experiments of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The KSTAR 5 GHz steady-state LHCD system is being designed to couple an input power of 2 MW for 300 s generated by four 5 GHz klystrons. For the development of this system, there are two critical issues. One is the development of a 5 GHz CW klystron for the RF source of the system. The other is the design of a steady-state LH launcher with active water cooling. In this paper, the current status of the development and design for the KSTAR steady-state LHCD system is described. For the LHCD system, aiming at a basic experimental study of 5 GHz LH wave propagation and operational experience with an LHCD system, the installation of an initial LHCD system with a capacity of 0.5 MW for 2 s is scheduled in 2010 using a 5 GHz prototype klystron and an un-cooled 1 MW launcher. The design and progress for the initial LHCD system are also presented.

Physics Basis of a Fusion Development Facility Utilizing the Tokamak Approach

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion’s energy applications. A concept of FDF based on the tokamak approach with conservative expressions of advanced physics and nonsuperconducting magnet technology is presented. It is envisioned to nominally provide 2 MW/m2 of neutron wall loading and operate continuously for up to 2 weeks as required for fusion nuclear component research and development. FDF will have tritium breeding capability with a goal of addressing the tritium self-sufficiency issue for fusion energy. A zero-dimensional system study using extrapolations of current physics and technology is used to optimize FDF for reasonable power consumption and moderate size. It projects a device that is between the DIII-D tokamak (major radius 1.8 m) [J. L. Luxon, Nucl. Fusion, Vol. 42, p. 614 (2002)] and the Joint European Torus (major radius 3 m) [P. H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion, Vol. 25, p. 1011 (1985)] in size, with an aspect ratio A of 3.5 and a fusion gain Q of 2 to 5. Theory-based stability and transport modeling is used to complement the system study and to address physics issues related to specific design points. It is demonstrated that the FDF magnetohydrodynamic stability limits can be readily met with conservative stabilizing conducting wall placement. Transport analysis using a drift-wave-based model with an edge boundary condition consistent with the pedestal stability limit indicates that the FDF confinement requirement can also be readily satisfied. A surprising finding is that the toroidal Alfvén eigenmodes are stabilized by strong ion Landau damping. Analysis of vertical stability control indicates that the basis configuration with an elongation κx [approximately] 2.35 can be controlled using a power supply technology similar to that used in DIII-D. Peak heat fluxes to the divertor are somewhat lower than those of ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Control. Fusion, Vol. 44, p. 519 (2002)], but FDF will operate with a higher duty factor.

Conceptual design of the FAST load assembly

Fusion advanced studies torus (FAST) is a proposal for a satellite facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact ( R 0 = 1.82 m, a = 0.64 m, triangularity δ = 0.4) machine able to investigate non-linear dynamics effects of alpha particle behaviours in burning plasmas [1,2,5]. The project is based on a dominant 30 MW of ion cyclotron resonance heating (ICRH), 6 MW of lower hybrid (LH) and 4 MW of electron cyclotron resonance heating (ECRH). FAST operates at a wide range [3,4] of parameters, e.g., in high performance H-mode ( B T up to 8.5 T; I P up to 8 MA) as well as in advanced Tokamak operation ( I P = 3 MA), and full non-inductive current scenario ( I P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170 s. To limit the TF magnet ripple ferromagnetic insert have been introduced inside the vacuum vessel (VV). Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI). Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plates material, and argon or neon as the injected impurities to mitigate the thermal loads.

Hollow current profile scenarios for advanced tokamak reactor operations

Advanced tokamak scenarios are a possible approach to boosting reactor performances. Such schemes usually trigger current holes, a particular magnetohydrodynamics equilibrium where no current or pressure gradients exist in the core of the plasma. While such equilibria have large bootstrap fractions, flat pressure profiles in the plasma core may not be optimal for a reactor. However, moderate modifications of the equilibrium current profile can lead to diamagnetism where most of the pressure gradient is now balanced by poloidal currents and the toroidal magnetic field. In this paper, we consider the properties of diamagnetic current holes, also called “dual equilibria,” and demonstrate that fusion throughput can be significantly increased in such scenarios. Their stability is investigated using the DCON code. Plasmas with a beta peak of 30% and an average beta of 6% are found stable to both fixed and free-boundary modes with toroidal mode numbers n=1–4, as well as Mercier and high-n ballooning modes. This is not surprising as these scenarios have a normal beta close to 3.

Commissioning the ICRF system and an ICRF assisted discharge cleaning at the KSTAR

Korean superconducting tokamak advanced research (KSTAR) is a national superconducting tokamak with the aim of a high beta operation based on advanced tokamak (AT) scenarios, and an ion cyclotron ranges of frequency (ICRF) heating is one of the essential tools to achieve this goal. The fabrication and high voltage (HV) test of the antenna and the matching system were finished in 2006 and the installation of the antenna, matching system and the transmitter at the KSTAR site was completed in 2007. Antenna conditioning was carried out to improve the HV holding condition of the antenna installed on the KSTAR and to check on the electro-magnetic (EM) interference with other equipments such as the superconducting magnet monitoring system and other machine and/or plasma diagnostic systems. The first KSTAR tokamak experimental campaign started by a vacuum pumping, a cryostat cooling and an ICRF system contributed to the successful tokamak shots through an ICRF assisted discharge cleaning of the vacuum vessel. In this paper, the installation processes of the ICRF system (with an emphasis on the quality assurance procedures of KSTAR), as well as the results from the first RF discharge experiment for the discharge cleaning and FWEH (fast wave electron heating) experiment for the KSTAR 1st experimental campaign are outlined.

Comprehensive control of resistive wall modes in DIII-D advanced tokamak plasmas

The resistive wall mode (RWM) and neoclassical tearing mode (NTM) have been simultaneously suppressed in the DIII-D for durations of over 2 s at beta values 20% above the no-wall limit with modest electron cyclotron current drive and very low plasma rotation. The achieved plasma rotation was significantly lower than reported previously. However, in this regime where stable operation is obtained, it is not unconditionally guaranteed. Various MHD activities, such as edge localized modes (ELMs) and fishbones, begin to couple to the RWM branch near the no-wall limit; feedback has been useful in improving the discharge stability to such perturbations. Simultaneous operation of slow dynamic error field correction and fast feedback suppressed the pile-up of ELM-induced RWM at a series of ELM events. This result implies that successful feedback operation requires not only direct feedback against unstable RWM but also careful control of MHD-induced RWM aftermath, which is the dynamical response to a small-uncorrected error field near the no-wall beta limit. These findings are extremely useful in defining the challenge of control of the RWM and NTM in the unexplored physics territory of burning plasmas in ITER.

Lower hybrid heating and current drive on the Alcator C-Mod tokamak

On the Alcator C-Mod tokamak, lower hybrid current drive (LHCD) is being used to modify the current profile with the aim of obtaining advanced tokamak (AT) performance in plasmas with parameters similar to those that would be required on ITER. To date, power levels in excess of 1 MW at a frequency of 4.6 GHz have been coupled into a variety of plasmas. Experiments have established that LHCD on C-Mod behaves globally as predicted by theory. Bulk current drive efficiencies, n20IlhR/Plh ∼ 0.25, inferred from magnetics and MSE are in line with theory. Quantitative comparisons between local measurements, MSE, ECE and hard x-ray bremsstrahlung, and theory/simulation using the GENRAY, TORIC-LH CQL3D and TSC-LSC codes have been performed. These comparisons have demonstrated the off-axis localization of the current drive, its magnitude and location dependence on the launched n∥ spectrum, and the use of LHCD during the current ramp to save volt-seconds and delay the peaking of the current profile. Broadening of the x-ray emission profile during ICRF heating indicates that the current drive location can be controlled by the electron temperature, as expected. In addition, an alteration in the plasma toroidal rotation profile during LHCD has been observed with a significant rotation in the counter-current direction. Notably, the rotation is accompanied by peaking of the density and temperature profiles on a current diffusion time scale inside of the half radius where the LH absorption is taking place.